ML021650244

License Amendment Request Pursuant to 10 CFR 50.90, Revision of Pressure/Temperature Limit Curves for Non-Nuclear Heatup/Cooldown, Core Critical Operation, & Pressure Testing for Reactor Coolant Systems

ML021650244
Person / Time
Site:	Perry
Issue date:	06/04/2002
From:	Kanda W FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
PY-CEI/NRR-2627L
Download: ML021650244 (167)

Text

Perry Nuclear Power Plant FENOC10 Center Road P.O. Box 97 FrstEnergyNuclear OperatingC-Pany Perry, Ohio 44081 June 4, 2002 PY-CEI/NRR-2627L United States Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 Perry Nuclear Power Plant Docket No. 50-440 License Amendment Request Pursuant to 10 CFR 50.90: Revision of Pressure/Temperature Limit Curves for Non-Nuclear Heatup/Cooldown, Core Critical Operation, and Pressure Testing for Reactor Coolant Systems; Including an Exemption Request Pursuant to 10 CFR 50.60(b)

Ladies and Gentlemen:

Nuclear Regulatory Commission (NRC) review and approval of an exemption and a license amendment for the Perry Nuclear Power Plant (PNPP) is requested. PNPP proposes changes to the Technical Specifications (TS) to revise the Reactor Coolant System (RCS)

Pressure/Temperature (P/T) limit curves specified in TS 3.4.11, "RCS Pressure and Temperature (P/T) Limits" for reactor non-critical heatup/cooldown, critical operation, and pressure testing for the RCS. The proposed amendment replaces the current RCS P/T limit curves contained in TS Figures 3.4.11-1 (a) through (f), with recalculated RCS P/T limit curves based, in part, on an alternative methodology.

The alternative methodology used to determine the new P/T limit curves has been endorsed by the American Society of Mechanical Engineers (ASME), but has not yet received formal approval for generic application by the NRC. The alternative methodology uses the ASME Boiler and Pressure Vessel (B&PV) Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1," in calculating the RCS P/T limits. Therefore, the use of this alternative methodology requires an exemption from the current requirements of 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation," pursuant to 10 CFR 50.60(b) and 10 CFR 50.12, "Specific Exemptions." The NRC has granted similar exemptions and approved the associated TS changes for the Clinton Power Station (ADAMS Accession Numbers ML003765298 and ML003765368) and the Riverbend Station (ADAMS Accession Numbers ML012280154 and ML012280403).

The procedures and methodology that were previously used to calculate the RCS P/T limit curves for PNPP were revised to recalculate the P/T limit curves, based, in part, on this ASME Code Case. During the recalculation of the P/T limit curves, an update to the PNPP neutron fluence calculations was also implemented. The methodology used to update the neutron fluence calculations is based upon the NRC-approved, "General Electric

Letter PY-CEI/NRR-2627L June 4, 2002 Page 2 of 2 Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation" (NEDC-32983P-A).

Therefore, the new P/T limit curves are based not only upon use of ASME Code Case N-640, but also upon use of revised neutron fluence values. provides a Summary and a Description of the proposed changes, a Safety Analysis, and an Environmental Consideration. Attachment 2 provides the Significant Hazards Consideration. Attachment 3 provides the marked-up annotated TS pages reflecting the requested changes. Attachment 4 provides the annotated TS Bases reflecting the proposed changes. Attachment 4 is for information only, since the TS Bases are not a formal part of the Technical Specifications. Attachment 5 provides the information justifying the Exemption Request. Attachment 6 provides the General Electric (GE) Nuclear Energy Report GE-NE-0000-0000-8763-01, "Pressure-Temperature Curves for First Energy Corporation, Using the K1c Methodology Perry Unit 1." GE-NE-0000-0000-8763-01 contains information that is proprietary to GE. Consistent with the proprietary information notice provided in the preface of the report, PNPP requests that the information provided by the report be withheld from public disclosure pursuant to 10 CFR 2.790(a)(4). Attachment 7 contains a non-proprietary version of the General Electric Report.

Application of the revised P/T limit curves is requested to be approved prior to March 1, 2003 in order to support Refueling Outage Nine (RFO9). Therefore, Nuclear Regulatory Commission review and approval is requested prior to this time. This request is considered a cost beneficial licensing change due to anticipated cost savings in outage duration.

Upon receipt of approval of the proposed license amendment, PNPP will submit, within 90 days, a revision to the Reactor Vessel Integrity Database (reference Generic Letter 92-01, "Reactor Vessel Structural Integrity", Revision 1, Supplement 1) to incorporate the data used in the development of the license amendment. If you have questions or require additional information, please contact Mr. Gregory A. Dunn, Manager - Regulatory Affairs, at (440) 280-5305.

Attachment 6 contains 2.790(a)(4)

VPROPRIETARY information. Upon removal of Attachment 6, the remainder of this letter and its William R. Kanda Attachments may be disclosed.

Vice President - Perry Attachments:

1. Summary and Description of the Proposed Technical Specification Changes

2. Significant Hazards Consideration

3. Technical Specification Pages Annotated with Proposed Change

4. Technical Specification Bases Pages Annotated with Proposed Change

5. Exemption Request

6. General Electric Report, GE-NE-0000-0000-8763-01, Proprietary Version

7. General Electric Report, GE-NE-0000-0000-8763-01a, Non-Proprietary Version cc: NRC Project Manager NRC Resident Inspector NRC Region III State of Ohio

I, William R. Kanda, hereby affirm that (1) I am Vice President - Perry, of the FirstEnergy Nuclear Operating Company, (2) I am duly authorized to execute and file this certification as the duly authorized agent for The Cleveland Electric Illuminating Company, Toledo Edison Company, Ohio Edison Company, and Pennsylvania Power Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief.

William R. Ida Subscribed to and affirmed before me, the $' day of _ v

/6 JANE E. MOTT Notary Public, State of Ohio My ComnrssIon Expires Feb. 20, 2005 (Recorded in Lake County)

Attachment 1 PY-CEI/NRR-2627L Page 1 of 6

SUMMARY

This license amendment and exemption request proposes to replace the current Technical Specification (TS) Pressure/Temperature (P/T) limit curves as specified in TS 3.4.11, "RCS Pressure and Temperature (P/T) Limits" [Figures 3.4.11-1 (a) through 3.4.11-1 (f)], with revised P/T limit curves. The revised curves will be for 22 and 32 Effective Full Power Years (EFPY).

The revised P/T limit curves are based, in part, on the application of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit Curves Section X1, Division 1." Code Case N-640 permits the use of an alternate fracture toughness curve, i.e., K1c instead of Kla, for the development of P/T curves. This method is an alternative to those currently approved by the NRC and recognized within the scope of 10 CFR 50.60. The use of an alternative method, therefore, requires an exemption from 10 CFR 50.60 requirements. This exemption is addressed in Attachment 5.

In developing the proposed P/T limit curves, the Perry Nuclear Power Plant (PNPP) neutron fluence calculations were also updated. These calculation updates were performed using the NRC-approved "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation" (General Electric Nuclear Energy Topical Report, NEDC-32983P-A, Revision 1).

Hence, the proposed PIT limit curves will also incorporate a revised Adjusted Reference Temperature (ART).

DESCRIPTION OF PROPOSED TECHNICAL SPECIFICATION CHANGE Technical Specification (TS) 3.4.11, "RCS Pressure and Temperature (P/T) Limits", replace the following figures:

Figure 3.4.11-1 (a), "Pressure Test Curve (Curve A)(Valid Up to 9 EFPY - Unit 11)",

Figure 3.4.11-1 (b), "Non-Nuclear Heatup/Cooldown (Curve B)(Valid Up to 9 EFPY Unit 1)",

Figure 3.4.11-1(c), "Core Critical Operation (Curve C)(Valid Up to 9 EFPY - Unit 1)",

Figure 3.4.11-1 (d), "Pressure Test Curve (Curve A)(Valid Up to 18 EFPY - Unit 1)",

Figure 3.4.11-1 (e), "Non-Nuclear Heatup/Cooldown (Curve B)(Valid Up to 18 EFPY Unit 1)", and Figure 3.4.11-1 (f), "Core Critical Operation (Curve C)(Valid Up to 18 EFPY - Unit 1)"

with figures:

Figure 3.4.11-1 (a), "Pressure Test Curve (Curve A)(Valid Up to 22 EFPY - Unit 11)",

Figure 3.4.11-1 (b), "Non-Nuclear Heatup/Cooldown (Curve B)(Valid Up to 22 EFPY Unit 1)",

Attachment 1 PY-CEI/N RR-2627L Page 2 of 6 Figure 3.4.11-1(c), "Core Critical Operation (Curve C)(Valid Up to 22 EFPY - Unit 1)",

Figure 3.4.11-1 (d), "Pressure Test Curve (Curve A)(Valid Up to 32 EFPY - Unit 1)",

Figure 3.4.11-1(e), "Non-Nuclear Heatup/Cooldown (Curve B)(Valid Up to 32 EFPY Unit 1)", and Figure 3.4.11-1(f), "Core Critical Operation (Curve C)(Valid Up to 32 EFPY - Unit 1)."

SAFETY ANALYSIS BACKGROUND 10 CFR 50, Appendix A, General Design Criteria (GDC) 31, "Fracture Prevention of Reactor Coolant Pressure Boundary", states that the reactor coolant pressure boundary shall be designed with sufficient margin to assure that when stressed under all modes of operation the boundary behaves in a nonbrittle manner. The GDC also states that design considerations include effects of irradiation on material properties. These requirements are re-iterated in 10 CFR 50.60, "Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation." The requirements of 10 CFR 50.60 are described within 10 CFR 50, Appendix G, "Fracture Toughness Requirements," and 10 CFR 50, Appendix H, "Reactor Vessel Material Surveillance Program Requirements."

10 CFR 50, Appendices G and H describe specific requirements for fracture toughness and reactor vessel material surveillance program requirements that must be considered in establishing Reactor Coolant System (RCS) Pressure/Temperature (P/T) limits. 10 CFR 50, Appendix G specifies that fracture toughness and testing requirements for reactor vessel material complies with the American Society of Mechanical Engineer (ASME) Boiler & Pressure Vessel (B&PV) Code. Appendix G also requires that the beltline material in the surveillance capsules be tested in accordance with the requirements of 10 CFR 50, Appendix H.

10 CFR 50, Appendix G endorses the ASME B&PV Code, Section Xl, Appendix G as providing a conservative method for developing the P/T limit curves. 10 CFR 50, Appendix G also requires the prediction of the effects of neutron irradiation on vessel embrittlement by calculating the Adjusted Reference Temperature [Adjusted Reference Temperature (ART) is the Reference Temperature (RTNDT) adjusted for the effects of neutron radiation] and the Charpy Upper Shelf Energy (USE). To predict these effects, Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials And Its Impact On Plant Operations," requires the methods described in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", be used.

Pursuant to 10 CFR 50, Appendix G, the materials used in the PNPP reactor vessel have been tested to determine their initial RTNDT and these values were used to develop the initial RCS P/T limit curves.

In February, 1996, a reactor vessel material surveillance capsule was removed from the PNPP Reactor Pressure Vessel (RPV). The materials were subsequently tested and analyzed. The methodology contained in Regulatory Guide 1.99, Revision 2 was used in the analysis of the materials. Based upon the results of this analysis, on August 28, 1997, a revision to the PNPP P/T curves was requested from the NRC. The NRC, as documented in

Attachment 1 PY-CEI/NRR-2627L Page 3 of 6 License Amendment 95 (TAC No. M99475), approved the revised P/T limits for PNPP. NRC approval was based on the conformance of the revised limits to the requirements of 10 CFR 50, Appendices G and H, and Regulatory Guide 1.99, Revision 2. Currently, TS Figures 3.4.11 -1(a) through (c) provides the predicted RCS P/T limit curves at the end of nine (9) EFPY for pressure testing (Curve A), for non-nuclear heatup/cooldown (Curve B),

and for core critical operations (Curve C), respectively. TS Figures 3.4.11-1 (d) through (f) provides identical information for 18 EFPY.

PNPP is at approximately 10.5 EFPY, and is using the 18 EFPY curves.

It should be noted, that the P/T limit curves implemented by License Amendment 95 were conservatively based upon neutron fluence values developed from the actual neutron flux the material specimens encountered instead of fluence values predicted by the analysis methodology used at the time the license amendment was developed.

ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGES In 2001, PNPP contracted the General Electric Company (GE) to recalculate the PNPP P/T limit curves. The recalculated (or proposed) P/T curves are based in part upon fluence calculations performed using a GE-specific methodology, which was used in the generation of the current P/T limit curves contained in TS Figures 3.4.11-1 (a) through (f). The recalculated (proposed) P/T curves also include improvements that have been made to the calculational methodology contained within the ASME B&PV Code,Section XI, Appendix G. The proposed P/T curves were calculated for 22 and 32 EFPY.

The methodology improvements were the application of the ASME B&PV Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1", and the use of the 1995 Edition with the 1996 Addenda of the ASME B&PV Code,Section XI, Appendix G. Code Case N-640 allows the use of the Kic fracture toughness curve rather than the Kia fracture toughness curve for use in the determination of the P/T limit curves. Use of the 1995 Edition with the 1996 Addenda of the ASME B&PV Code,Section XI, is required by 10 CFR 50, Appendix G, § IV.A.1.a, and 10 CFR 50.55a(b)(2).

Due to the desire to implement the above-mentioned code case, an opportunity to update the PNPP neutron fluence calculations became available. Therefore, GE was also tasked to update the fluence calculations. The NRC-approved "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation" (NEDC-32983P-A) was used to perform the fluence update. Hence, the proposed P/T limit curves are not only based upon use of revised ASME B&PV Code methodology, but also upon the incorporation of the effects from updated fluence values.

Application of ASME Code Case The proposed P/T limits were developed based on the methodology specified in ASME B&PV Code Section XI, Appendix G, as modified by ASME B&PV Nuclear Code Case N-640. Code Case N-640 allows the use of alternate material fracture toughness when determining minimum vessel temperatures. Specifically, Kic values defined in ASME B&PV Code,Section XI, Appendix A, Figure A-4200-1 will be used instead of the K1a values defined in ASME B&PV Code, Section Xl, Appendix G, Figure G-2210-1.

Attachment 1 PY-CEI/NRR-2627L Page 4 of 6 Use of the K1, curve in determining the lower bound fracture toughness in the development of P/T limit curves is more technically correct than the Kia curve. The K1c curve models the slow heatup and cooldown processes that a Reactor Pressure Vessel (RPV) normally undergoes. These slow heatup and cooldown limits are enforced through the use of the PNPP Technical Specification 3.4.11, "RCS Pressure and Temperature (P/T) Limits."

Surveillance Requirement 3.4.11.1 states that heatup and cooldown rates are to be < 100- F in any one hour period.

Use of this approach is justified by the initial conservatism of the Kla curve when the curve was incorporated into the ASME B&PV Code in 1974. This initial conservatism was necessary due to limited knowledge of RPV material fracture toughness. Since 1974, additional knowledge has been gained about the fracture toughness of RPV materials and their fracture response to applied loads. The additional knowledge demonstrates that the fracture toughness provided by the K1a curve is well beyond the margin of safety required to protect against potential RPV failure, and that the K1c fracture toughness curve provides an adequate margin of safety to protect against potential RPV failure.

The acceptability of and the technical basis for use of Code Case N-640 is described in "Technical Basis for Revised P-T Limit Curve Methodology," by W. H. Bamford (Westinghouse Electric), S. N. Malik (NRC), et. al. This methodology was presented at the 2000 ASME Pressure Vessels and Piping Conference. Simply, the revised methodology removes excess conservatism in the current ASME, Appendix G approach. Performance of leak tests at artificially high temperatures could impact test personnel safety, challenge operators with maintaining a high temperature in a limited operating band, and decreases availability of plant systems, including the Residual Heat Removal System, due to the longer RPV heatup and test time.

Notwithstanding the above-described changes to the methodology used to calculate the proposed P/T limit curves, the modified methodology satisfies the guidance contained in the 1995 Edition, 1996 Addenda of the ASME B&PV Code, Section Xl, Appendix G. Hence, the intent of the guidance contained in 10 CFR 50, Appendices G and H is also satisfied.

The NRC has found the application of this code case acceptable, a number of licensees have requested the use of Code Case N-640 and their requests have been approved.

[Reference - Clinton Power Station (ADAMS Accession Numbers ML003765298 and ML003765368) and the Riverbend Station (ADAMS Accession Numbers ML012280154 and ML012280403)]. It should be noted that the NRC is in the process of providing approval of this Code Case for generic use by including the case into Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability ASME Section XI, Division 1" (Reference Draft Regulatory Guide DG-1 091 - 66 Federal Register 67335).

Based upon the technical basis provided in "Technical Basis for Revised P-T Limit Curve Methodology", and continued compliance with 10 CFR 50, Appendices G and H, the PNPP staff has determined that the proposed P/T limit curves maintain an adequate margin of safety for brittle fracture.

Attachment 1 PY-CEI/NRR-2627L Page 5 of 6 Application of the ASME B&PV Code 1995 Edition with 1996 Addenda The proposed P/T limits were developed using Section XI, Appendix G of the 1995 Edition with the 1996 Addenda of the ASME B&PV Code. This code edition and addenda incorporated revised stress intensity factors into the Appendix G methodology, which is used to develop the actual PiT limit curves. The revised stress intensity factors are based upon the re-orientation of the postulated defect normal to the direction of maximum stress. This code edition with addenda has been approved for use by the NRC as documented in 10 CFR 50.55a(b)(2).

Update of Fluence Calculations 10 CFR 50, Appendix A, GDC 31 and 10 CFR 50, Appendix G requires the prediction of the effects of neutron irradiation on vessel embrittlement. To predict these effects, the NRC requires the methods described in Regulatory Guide 1.99, Revision 2, be used. The regulatory guide requires factors to be calculated to account for the effects of neutron embrittlement. The factors are the Adjusted Reference Temperature (ART) and the Charpy Upper Shelf Energy (USE). Ifthe ART and the USE satisfy the limits contained in the regulatory guide and 10 CFR 50, Appendix G, then the vessel materials provide adequate margin against brittle fracture.

One of the key components used in the calculations of both the ART and the USE, is the RPV fluence. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", provides guidance for the calculation of RPV fluence. The fluence values calculated using the methodology described in this Regulatory Guide satisfy the requirements of 10 CFR 50, Appendix G and Regulatory Guide 1.99.

The General Electric (GE), "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation" (NEDC-32983P-A), methodology was used to determine the PNPP RPV fluence for use in the development of the proposed P/T curves. The GE methodology is based upon a two-dimensional, discrete ordinates code, which also employs Monte-Carlo techniques to calculate the neutron fluence. Regulatory Guide 1.190 states that the discrete ordinate and Monte-Carlo methodologies are acceptable methodologies for the determination of neutron fluence. The GE methodology has been approved by the NRC (ADAMS Accession Number ML012400381). As stated in the NRC Safety Evaluation for the GE methodology, the methodology is consistent with Regulatory Guide 1.190.

Using the results of the new fluence calculations, revised values of the ART were calculated.

Comparing the revised ART values with the values calculated to support License Amendment 95, the limiting beltline component (Weld Heat 627260) remained the same, with the ART value being slightly lower than its previous value. This was not unexpected due to the use of an improved fluence methodology. Nevertheless, the revised ART value for the limiting beltline component is well below the regulatory limit.

Using the results of the new fluence calculations, revised values of the USE were calculated. Comparing the revised USE values with the values calculated to support License Amendment 95, the limiting component (Plate 22-1-1, Heat C2557-1) and its associated USE remained the same.

Attachment 1 PY-CEI/NRR-2627L Page 6 of 6 Since the predicted lowest upper shelf energy at 32 EFPY was greater than the minimum of 50 ft-lbs required by 10 CFR 50, Appendix G and the ART for the limiting material was less than the 2000 F limit required by Regulatory Guide 1.99, Revision 2, the integrity of the RCS has been maintained.

The calculations used in the development of the proposed P/T curves are contained in . Since the GE methodology and PNPP-specific application of the methodology are considered proprietary, Attachment 7 contains a non-proprietary version of the calculations contained within Attachment 6.

CONCLUSION NRC regulations require that P/T limit curves provide an adequate margin of safety to the conditions at which brittle fracture may occur. These requirements are set forth in GDC 31 and 10 CFR 50, Appendices G and H. Regulatory Guides 1.99 and 1.190 provide guidance for the compliance with the requirements of the GDC and the appendices. The appendices reference the requirements and guidance of ASME B&PV Code, Section Xl, Appendix G for the development of P/T limit curves. The methodologies described within the regulatory guides and the ASME Code will provide P/T limit curves with the requisite margin against brittle fracture. The proposed P/T limit curves are based on these methodologies as modified by application of ASME Code Case N-640.

Although the code case proposes a change to a requirement contained in ASME,Section XI, Appendix G, the alternative allowed by Code Case N-640 is based upon industry experience gained since the inception of 10 CFR 50, Appendix G. The more appropriate assumptions and provisions allowed by the code case maintain a margin of safety that is consistent with the intent of 10 CFR 50, Appendices G and H.

This proposed license amendment and exemption request are similar in scope to license amendments and exemptions already granted to the Clinton Power Station (ADAMS Accession Numbers ML003765298 and ML003765368) and the River Bend Station (ADAMS Accession Numbers ML012280154 and ML012280403).

ENVIRONMENTAL CONSIDERATION The proposed Technical Specification change request was evaluated against the criteria of 10 CFR 51.22 for environmental considerations. The proposed change does not significantly increase individual or cumulative occupational radiation exposures, does not significantly change the types or significantly increase the amounts of effluents that may be released off-site, and as discussed in Attachment 2, does not involve a significant hazards consideration. Based upon the foregoing, it has been concluded that the proposed Technical Specification change meets the criteria given in 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an Environmental Impact Statement.

Attachment 2 PY-CEI/NRR-2627L Page 1 of 3 SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that a request for amendment involves no significant hazards considerations are included in the Nuclear Regulatory Commission's Regulation, 10 CFR 50.92, which states that the operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any previously evaluated; or (3) involve a significant reduction in a margin of safety.

The proposed amendment is requesting Nuclear Regulatory Commission review and approval of changes to the Perry Nuclear Power Plant (PNPP) Technical Specifications to incorporate revised Reactor Coolant System (RCS) Pressure/Temperature (P/T) limit curves.

The proposed amendment has been reviewed with respect to these three factors and it has been determined that the proposed change does not involve a significant hazard because:

1. The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed P/T limit curves are based upon the use of an alternate material fracture toughness curve and the use of an NRC-approved methodology for calculation of neutron fluence. The proposed RCS P/T limit curves are valid through 22 Effective Full-Power Years (EFPY) and 32 EFPY.

The American Society of Mechanical Engineers (ASME) Boiler & Pressure Vessel (B&PV) Code Case N-640 permits the use of KIc as defined in ASME B&PV Code,Section XI, Appendix A, Figure A-4200-1 instead of Kia as defined in ASME B&PV Code,Section XI, Appendix G, Figure G-221 0-1. The use of the Kic curve in determining the lower bound fracture toughness in the development of P/T limit curves is more technically correct than the Kia curve. The KIc curve models the slow heatup and cooldown processes that a Reactor Pressure Vessel (RPV) normally undergoes. These slow heatup and cooldown limits are enforced through the use of the PNPP Technical Specification 3.4.11, "RCS Pressure and Temperature (P/T)

Limits." Surveillance Requirement 3.4.11.1 states that heatup and cooldown rates will be < 100-° F in any one hour period. The use of the K1c curve is applicable to PNPP and is in consistent with the ASME B&PV. Therefore, the use of KI, will provide an adequate margin of safety to protect against potential RPV failure.

NRC regulations require the vessel material transition temperature be adjusted to account for the effects of neutron radiation. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence", provides a methodology for calculating the neutron fluence, while Regulatory Guide 1.99, "Radiation Embrittlement of Reactor Vessel Materials", provides the guidance for calculating the adjusted transition temperature using the fluence factor. The methodologies satisfy the requirements of 10 CFR 50, Appendices G and H, and General Design Criteria 31, "Fracture Prevention of Reactor Coolant Pressure

Attachment 2 PY-CEI/NRR-2627L Page 2 of 3 Boundary." The methodologies used to develop the proposed P/T limit curves satisfy the requirements of the regulations. The predicted lowest upper shelf energy at 32 EFPY was greater than the minimum of 50 ft-lbs required by 10 CFR 50, Appendix G. The adjusted reference temperature for the limiting material was less than the 2000 F limit required by Regulatory Guide 1.99, Revision 2. Therefore, the integrity of the RCS has been maintained. As such, the proposed curves ensure that adequate reactor vessel safety margins against nonductile failure exist during normal operation, anticipated operational occurrences, and hydrostatic testing. There are no plant modifications associated with these changes. Thus, the proposed changes do not involve a significant increase in the probability of occurrence of an accident previously evaluated.

The proposed changes do not adversely affect the integrity of the reactor vessel.

Hence, the function of the reactor vessel to act as a radiological barrier during an accident is not affected. Therefore, the proposed changes do not involve a significant increase in the consequences of an accident previously evaluated.

2. The proposed change would not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed P/T limit curves are based upon the use of an alternate material fracture toughness curve and the use of an NRC-approved methodology for calculation of neutron fluence.

The ASME B&PV Code Case N-640 permits the use of the K1c curve in determining the lower bound fracture toughness in the development of P/T limit curves. The K1c curve models the slow heatup and cooldown processes that a RPV normally undergoes. These slow heatup and cooldown limits are enforced through the use of the PNPP Technical Specifications. Therefore, the use of K1c will provide an adequate margin of safety to protect against potential RPV failure.

NRC regulations require the vessel material transition temperature be adjusted to account for the effects of neutron radiation. The methodologies used to develop the proposed P/T limit curves satisfy the requirements of the regulations. The predicted lowest upper shelf energy at 32 EFPY was greater than the minimum of 50 ft-lbs required by 10 CFR 50, Appendix G. The adjusted reference temperature for the limiting material was less than the 2000 F limit required by Regulatory Guide 1.99, Revision 2. Therefore, the integrity of the RCS has been maintained. As such, the proposed curves ensure that adequate reactor vessel safety margins against nonductile failure exist during normal operation, anticipated operational occurrences, and hydrostatic testing.

There are no plant modifications associated with these changes.

The proposed changes to the P/T limit curves do not affect the assumed accident performance of any structure, system, or component previously evaluated. The proposed changes do not introduce any new modes of system operation or failure mechanisms. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

Attachment 2 PY-CEI/NRR-2627L Page 3 of 3

3. The proposed change will not involve a significant reduction in the margin of safety.

NRC regulations require that P/T limits provide an adequate margin of safety to the conditions at which brittle fracture may occur. These requirements are set forth in 10 CFR 50, Appendix A, General Design Criteria (GCD) 31, and 10 CFR 50, Appendices G and H. Regulatory Guides 1.99 and 1.190 provide guidance for the compliance of GDC 31 and Appendices G and H. The appendices reference the requirements and guidance of ASME B&PV Code, Section Xl, Appendix G for the development of P/T limit curves. The methodologies described within the regulatory guides and the ASME Code will provide P/T limit curves with the requisite margin against brittle fracture. The proposed P/T limit curves are based on these methodologies as modified by application of ASME Code Case N-640.

Although the code case proposes a change to a requirement contained in ASME,Section XI, Appendix G, the alternative allowed by Code Case N-640 is based upon industry experience gained since the inception of 10 CFR 50, Appendix G. The more appropriate assumptions and provisions allowed by the code case maintain a margin of safety that is consistent with the intent of 10 CFR 50, Appendices G and H.

Therefore, the proposed changes do not involve a significant reduction in the margin of safety.

Attachment 3 PY-CEI/NRR-2627L Page 1 of 12 a.

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w LU MINIMUM REACTOR VESSEL I Figure 3.4.11-1(a): Pressure Test Curve PERRY - UNIT 1 3.4-31 Amendment No. 95

Attachment 3 PY-CEI/NRR-2627L Page 2 of 12 f 1 Pa e2 RC PTFLimits 3.4.11 INITIAL RTndt VALUES ARE 10F FOR BELTLINE.

-20°F FOR UPPER VESSEL. AND 10F FOR BOTTOM HEAD HEATUP/COOLDOWN S

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Attachment 3 PY-CEI/NRR-2627L Page v 3 of 12 S P/T Limits 3.4.11 INITIAL RTndt VALUES ARE 10"F FOR BELTLINE.

-20°F FOR UPPER VESSEL. AND 10"F FOR BOTTOM HEAD HEATUPICOOLDOWN RATE 100°FIHR BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (-F)

UPTO9 30.3 PERRY - UNIT 1 3.4-31 b Amendment No. 95 I .

• CS P/T Limits 3.4.11 nt ARE -30"F FOR BELTLINE

-20"F FOR UPPER VESSEL, AND 10.F FOR BOTTOM HEAD CHEATUP/COOLDOWN RATE 20OF/HR ADJUSTED CURVES BELTUNEAS SHOWN:

EFPY SHIFT (F)

UP TO 18 91.9 PERRY - UNIT 1 Amendment 3.4-31 c No. 95 1

Attachment 3 PY-CEI/NRR-2627L Page 5 of 12 RCS P/T Limits 3.4.11

.R-nd t VA LUES RE -30"F FOR BELTUNG,

-20"F FOR UPPER VESSEL, AND 10°F FOR BOTTOM HEAD HEATUP/COOLDOWN RATE 1001FIHR BELTLINE CURVES ADJUSTED AS SHOWN:

EFPY SHIFT (-F)

BELT tE LIMITS S*BOTTO EA LIMITS 400 URE (-F) iP to IS EFPY *Unit 1)

PERRY - UNIT 1 3.4-31 d Amendment No. 95

Attachment 3 PY-CEI/NRR-2627L Page 6 of 12 1400 1200 CL 1000 1 0

.j o 800 iw BELTLINE CURVES

> ADJUSTED AS SHOWN:

0 EFPY SHIFT (F) tie" 60UP TO 18 91.9 LuI S'Minimum Criticality Temperature " BELTLIN 70'F NON-BELINE 200 . LIMITS 0

0 100 200 300 400 MIN UM REACTOR VESSEL METAL TEMPERATURE (-F)

Figure 3.4.11-1 f): Core Critical Operation (Curve C)(Valid Up to 18 EFPY - Unit 1)

PERRY - UNIT 1 3.4-31 e Amendment No. 95

Attachment 3 PY-CEI/NRR-2627L Page 7 of 12 1400 1300 HEATUP/COOLDOW]

RATE 20°F/HR 1200 1100 1000 a.

LU.

o 900 0

I

,J (A

,, 800 0 700 Iz 600 U.'

500 U.'

400 w

IL 300 UPPER VESSEL 200 AND BELTLINE LIMITS

- ------ BOTTOM HEAD 100 CURVE 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(*F)

Figure 3.4.11-1(a): Pressure Test Curve (Curve A) (Valid Up to 22 EFPY - Unit 1)

Attachment 3 PY-CEI/NRR-2627L Page 8 of 12 1400 1300 1200 HEATUP/COOLDOWN 1100 I RATE 100°F/HR a.

1000

'U IL (1.

900 0

I

-J ILu 800

'U 0 700 I

z 600

_1 500 0

0 400 a.

300

-UPPER VESSEL 200 AND BELTLINE LIMITS 100 ------.BOTTOM HEAD CURVE 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 3.4.11-1(b): Non-Nuclear Heatup/Cooldown (Curve B) (Valid Up to 22 EFPY - Unit 1)

Attachment 3 PY-CEI/NRR-2627L Page 9 of 12 1400 1300 1200 HEATU P/COOLDOWN RATE 100°F/HR 1100 Is

" 1000 900 F

w 0 S80 o 700 Z

0 S60 S500 w

0 S40 w

Di 300 200 UPPER VESSEL AND BELTLINE LIMITS 100 ------.BOTTOM HEAD CURVE 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 3.4.11-1(c): Core Critical Operation (Curve C) (Valid Up to 22 EFPY - Unit 1)

Attachment 3 PY-CEI/NRR-2627L Page 10 of 12 1400 1300 HEATUP/COOLDOWN 1200 RATE 20°F/HR 1100 CL

0. 1000 x

DI

0. 900 Co

-J 800 0 700 I

w 600 Co Co I

500 0.

400 300 UPPER VESSEL 200 AND BELTLINE LIMITS

- ------ BOTTOM HEAD 100 CURVE 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 3.4.11-1(d): Pressure Test Curve (Curve A) (Valid Up to 32 EFPY - Unit 1)

Attachment 3 PY-CEI/NRR-2627L Page 11 of 12 1400 HEATUP/COOLDOWN 1300 RATE 100°F/HR 1200 1100 S1000 4

0 900

_j

-j S800 vi o 700 Uj S600 S5 0 0 LU S4 0 0 U.'

D.

300 UPPER VESSEL 200 AND BELTLINE LIMITS 100 ------ BOTTOM HEAD CURVE 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 3.4.11-1(e): Non-Nuclear Heatup/Cooldown (Curve B) (Valid Up to 32 EFPY - Unit 1)

Attachment 3 PY-CEI/NRR-2627L Page 12 of 12 1400 1300 HEATUP/COOLDOWN 1200 RATE 100°F/HR 1100 0 1000 I.

U.

Ja.

0 900 I

-J 0

800 Co I

700 I-600 U.'

IL 500 IU.

Co Co 400 U.'

300

-UPPER VESSEL 200 AND BELTLINE LIMITS 100 - ------ BOTTOM HEAD CURVE 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 3.4.11-1(f): Core Critical Operation (Curve C) (Valid Up to 32 EFPY - Unit 1)

Attachment 4 PY-CEI/NRR-2627L RCS P/T Limits Page 1 of 2 B 3.411 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.11 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) shutdown (cooldown) operations, power transients, and and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.

Figure 3.4.11-1 contains P/T limit curves for heatup; cooldown. and inservice leak and hydrostatic testing.

The up curve 'ides limits for both heatup and "cri 'icaly Cures are provided which are valid for up to 15 E a d 'J'81 EFPYY

(/Timi cur e defines an acceptable region for normal operation.> e ual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of reactor coolant pressure boundary (RCPB). the The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel.

10 CFR 50, Appendix G (Ref. 1). requires the of P/T limits for material fracture toughness establishment requirements of the RCPB materials. Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the:use of the American Society of Mechanical Engineers (ASME) Code,Section III, Appendix G

(Ref. 2).

The actual shift in the RTnDT of the vessel material will established periodically by removing and evaluating the be irradiated reactor vessel material specimens, in accordance with ASTM E 185-82 (Ref. 3) and 10 CFR 50. Appendix H (continued)

PERRY - UNIT 1 .- B 3-4-54 Revision No. 3

Attachment 4 PY-CEI/NRR-2627L RCS P/T Limits Page 2 of 2 B 3.4.11 BASES SURVEILLANCE SR 3.4.11.10 (continued)

REQUIREMENTS material specimens, in accordance with ASTM E 185-82 (Ref. 3) and 10 CFR 50, Appendix H (Ref. 4). The operating P/T limit curves in Figure 3.4.11-1 will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.

REFERENCES 1. 10 CFR 50, Appendix G.

2. ASME, Boiler and Pressure Vessel Code,Section III, Appendix G.

3. ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests For Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

6. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

7. -,, e DRi Af' e'--

Fretr rements For BuiWe."

Deccmbc r_,4_ll-7..8.

19

88. USAR, Section 15.4.4.

/9. GE Services Information Letter, SIL No. 517 Supplement 1. 'Analysis Basis for Idle Recirculation Loop Startup."

/ - ,V*V 5 0 5-3 0 .5 PERRY - UNIT 1 B 3.4-63 Revision No. 3

Attachment 5 PY-CEI/NRR-2627L Page 1 of 4 EXEMPTION REQUEST The Pressure/Temperature (P/T) limits proposed by the Perry Nuclear Power Plant (PNPP) are calculated using an alternative method to that described in 10 CFR 50, Appendices G and H.

The alternative method, in part, is based upon the use of an American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Case. Specifically, Code Case N-640, "Alternative Reference Fracture Toughness for Development of P-T Limit CurvesSection XI, Division 1", will be used in calculating the proposed PNPP Reactor Coolant System (RCS) PIT limits. Since this code case has not yet received formal approval from the NRC for generic application, the use of the alternative method requires an exemption from the current requirements of 10 CFR 50.60, "Acceptance Criteria", which implements 10 CFR 50, Appendices G and H.

Pursuant to 10 CFR 50.12, the NRC may grant an exemption from requirements contained in 10 CFR 50 (in this case 10 CFR 50.60) provided the following four conditions are satisfied:

1. The requested exemption is authorized by law,

2. The requested exemption does not present an undue risk to the public health and safety,

3. The requested exemption will not endanger the common defense and security, and

4. Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.60.

Exemptions to Code Case N-640 have been granted by the NRC to numerous nuclear facilities including the Clinton Power Station (ADAMS Accession Number ML003765298) and the River Bend Station (ADAMS Accession Number ML012280154). Additionally, the NRC is processing the generic approval of this Code Case for inclusion into Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability ASME Section XI, Division 1"(Reference Draft Regulatory Guide DG-1 091 - 66 Federal Register 67335).

ASME B&PV Code Case N-640 10 CFR 50.12(a) Requirements The requested exemption to allow use of ASME B&PV Code Case N-640 in conjunction with ASME B&PV Code XI, Appendix G to determine the pressure-temperature limits for the reactor pressure vessel meets the criteria of 10 CFR 50.12 as discussed below.

1. The requested exemption is authorized by law:

10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendices G and H when an exemption is granted by the Commission under 10 CFR 50.12.

Attachment 5 PY-CEI/NRR-2627L Page 2 of 4

2. The requested exemption does not present an undue risk to the public health and safety:

The proposed P/T limits for the PNPP rely in part on the requested exemption. In accordance with Code Case N-640, the proposed P/T limits have been developed using the Klcfracture toughness curve shown on ASME B&PV Code,Section XI, Appendix A, Figure A-4200-1, in lieu of the Kia fracture toughness curve of ASME B&PV Code,Section XI, Appendix G, Figure G-2210-1. This curve is used as the lower bound for fracture toughness. Except for the changes associated with Code Case N-640, the other margins involved with the ASME B&PV Code,Section XI, Appendix G process of determining P/T limit curves remain unchanged.

Use of the Kic curve in determining the lower bound fracture toughness in the development of P/T operating limits curve is more technically correct than the Kia curve.

The K1, curve models the slow heatup and cooldown processes that a RPV undergoes.

These slow heatup and cooldown limits are enforced through the use of Technical Specification 3.4.11, "RCS Pressure and Temperature (P/T) Limits." Surveillance Requirement 3.4.11.1 states that heatup and cooldown rates are to be < 100-° F in any one hour period.

Use of this approach is justified by the initial conservatism of the Kia curve when the curve was incorporated into the ASME B&PV Code in 1974. This initial conservatism was necessary due to limited knowledge of RPV material fracture toughness. Since 1974, additional knowledge has been gained about the fracture toughness of RPV materials and their fracture response to applied loads. The additional knowledge demonstrates that the fracture toughness provided by the KIa curve is well beyond the margin of safety required to protect against potential RPV failure, and that the K1c fracture toughness curve provides an adequate margin of safety to protect against potential RPV failure.

The use of P/T curves based on the K1, fracture toughness limits will enhance overall plant safety by widening the P/T operating window especially in the region of low temperature operations. Safety benefits that would be realized during the pressure test include a reduction in the challenges to operators in maintaining a high temperature in a limited operating band, personnel safety while conducting inspections in primary containment at elevated temperatures, and increased availability of plant systems, including the Residual Heat Removal System, due to reduction of the heatup and test time.

Based on the above, this exemption does not present an undue risk to the public health and safety.

3. The requested exemption will not endanger the common defense and security:

This exemption request concerns the revision of operating and test limits for the PNPP commercial power reactor in accordance with industry-proposed guidance and has no impact on common defense and security. Therefore, the common defense and security are not endangered by approval of this exemption request.

Attachment 5 PY-CEI/NRR-2627L Page 3 of 4

4. Special circumstances are present which necessitate the request for an exemption to the regulations of 10 CFR 50.60:

In accordance with 10 CFR 50.12(a)(2), the NRC will consider granting an exemption to the regulations if "special circumstances" are present. The regulation provides six criterion which licensees can use to provide the basis for the "special circumstance" provision of the regulation. The following three criterion are applicable to this exemption request:

"(a)(2)(ii) - Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule; or (a)(2)(iii) - Compliance would result in undue hardship or other costs that are significantly in excess of those contemplated when the regulation was adopted, or that are significantly in excess of those incurred by others similarly situated; or (a)(2)(v) - The exemption would provide only temporary relief from the applicable regulation and the licensee or applicant has made good faith efforts to comply with the regulations."

Each of the above requirements is addressed below.

10 CFR 50.12(a)(2)(ii):

The ASME B&PV Code,Section XI, Appendix G, provides procedures for determining allowable loading on the RPV and is approved for that purpose by 10 CFR 50, Appendix G. Application of these procedures in the determination of P/T operating and pressure test limit curves satisfy the underlying requirement that the reactor coolant pressure boundary be operated in a regime having sufficient margin to ensure, that when stressed, the RPV boundary will behave in a non-brittle manner. Use of these procedures provides assurance that the probability of a rapidly propagating fracture will be minimized. Hence, P/T operating and test limit curves developed using these procedures ensures that adequate margin exists considering the uncertainties in determining the effects of irradiation on material properties.

The ASME B&PV Code,Section XI, Appendix G, procedure was conservatively developed based on the level of knowledge existing in 1974 concerning RPV materials and the estimated effects of operation. Since 1974, the level of knowledge about these topics has been greatly expanded. This increased knowledge permits relaxation of the ASME B&PV Code, Section Xl, Appendix G, requirements via application of the ASME B&PV Code Case N-640, while maintaining the underlying purpose of the ASME B&PV Code and the NRC regulations to ensure an acceptable margin of safety.

10 CFR 50.12(a)(2)(iii):

The reactor coolant system pressure-temperature operating window is defined by the P/T operating and test limit curves developed in accordance with the ASME B&PV Code,Section XI, Appendix G procedure. Continued operation of the PNPP, with these P/T curves without the relief provided by ASME B&PV Code Case N-640 would unnecessarily restrict the pressure-temperature operating band. This restriction

Attachment 5 PY-CEI/NRR-2627L Page 4 of 4 challenges the operations staff during pressure tests to maintain a high temperature within a limited operating band. It also subjects inspection personnel to increased safety hazards while conducting inspections of systems at elevated temperatures.

This constitutes an unnecessary burden that can be alleviated by the application of ASME B&PV Code Case N-640 in the development of the proposed P/T curves.

Implementation of the proposed P/T curves as allowed by ASME B&PV Code Case N-640 does not significantly reduce the margin of safety below that established by the original requirement.

10 CFR 50.12(a)(2)(yv):

The requested exemption provides only temporary relief, since the PNPP staff anticipates that the provisions of Code Case N-640 will be incorporated into (or reconciled with) the requirements of 10 CFR 50, Appendix G, based on ongoing industry efforts to do so. NRC approval of the Code Case is pending, but additional action may be required to allow use of the Code Case without requiring an exemption to 10 CFR 50, Appendix G. The estimated time for such actions to be completed is unknown, and therefore, the effective period of time that the exemption would be effective is indefinite.

Summary for ASME B&PV Code Case N-640 Compliance with the specified requirement of 10 CFR 50.60(a) would result in hardship and unusual difficulty without a compensating increase in the level of quality and safety. ASME B&PV Code Case N-640 allows a reduction in the lower bound fracture toughness used in ASME B&PV Code, Section Xl, Appendix G, in the determination of RCS P-T limits. This proposed alternative is acceptable because the ASME B&PV Code Case maintains the relative margin of safety commensurate with that which existed at the time ASME B&PV Code,Section XI, Appendix G, was approved in 1974. Therefore, application of ASME B&PV Code Case N-640 for the PNPP will ensure an acceptable margin of safety and does not present an undue risk to the public health and safety.

Attachment 7 PY-CEI/NRR-2627L GE NuclearEnergy Engineering and Technology GE-NE-0000-0000-8763-01 a General Electric Company Revision 0 175 Curtner Avenue, Class I San Jose, CA 95125 April 2002 Pressure-Temperature Curves For FirstEnergy Corporation, Using the Kc Methodology Perry Unit I Prepared by: _A08Ae6&W- -Par Matthew O'Connor, Mechanical Engineer Structural Assessment and Mitigation Verified by:

L.J. lilly, Senior Engineer Structural Assesaffignd'4Jitiaation Approved by:

Betty Brdnlund, Principal Engineer Structural Assessment and Mitigation

GENE 0000-0000-8763-01a Revision 0 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this document are contained in the contract between FirstEnergyCorporationand GE, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than FirstEnergy Corporation, or for any purpose other than that for which it is intended is not authorized. With respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

GENE 0000-0000-8763-O1 a Revision 0 Executive Summary This report provides the pressure-temperature curves (P-T curves) developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beitline limits and irradiation embrittlement effects in the beitline. The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in November 1996 [1]. Several improvements were made to the P-T curve methodology; the improvements include, but are not limited to the following: 1) The incorporation of ASME Code Case N-640. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress. ASME Code Case N-640 allows the use of K1c rather than Kia to determine T-RTNDT. Descriptions of other improvements are included in the P-T curve methodology section.

Conclusions The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:

"* Closure flange region (Region A)

"* Core beltline region (Region B)

"* Upper vessel (Regions A & B)

"* Lower vessel (Regions B & C)

- ii -

GENE 0000-0000-8763-01a Revision 0 For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr or less for which the curves are applicable. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in this report. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20°F/hr or less must be maintained at all times. Additionally, P-T curves for the core not critical and core critical conditions, with a 200°F/hr heatup/cooldown temperature rate are provided.

The P-T curves apply for both heatup/cooldown and for both the 1/4T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, Kir, at 1/4T to be less than that at 3/4T for a given metal temperature.

Composite P-T curves were generated for each of the Pressure Test, Core Not Critical and Core Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate bottom head, beltline, upper vessel and closure assembly P-T limits. Separate P-T curves were developed for the upper vessel, beltline and bottom head for the Pressure Test and Core Not Critical conditions. In addition, beltline limits are provided for 22 EFPY for the Pressure Test and Core Not Critical conditions, and a composite curve for 22 EFPY is provided for the Core Critical condition.

- iii -

GENE 0000-0000-8763-01a Revision 0 Table of Contents Page 1.0 Introduction .......................................................................................... 1 2.0 Scope of the Analysis ...................................................................... 3 3.0 Analysis Assumptions ................................................................... 5 4.0 Analysis ........................................................................................... 6 4.1 Initial Reference Temperature ................................................. 6 4.2 Adjusted Reference Temperature for Beltline ..................... 11 4.3 Pressure-Temperature Curve Methodology .......................... 15 5.0 Conclusions and Recommendations ........................................... 42 6.0 References ..................................................................................... 57

- iv -

GENE 0000-0000-8763-01a Revision 0 Table of Appendices Appendix A Description Of Discontinuities Appendix B Tabulation Of P-T Curves Appendix C Operating And Temperature Monitoring Requirements Appendix D GE SIL 430 Appendix E Determination of Upper Shelf Energy Appendix F Pressure-Temperature Curves (200°F/hr HeatuplCooldown)

Appendix G Determination of Beltline Region and the Impact on Fracture Toughness Appendix H Determination of Peak Vessel Fluence

GENE 0000-0000-8763-01 a Revision 0 Table of Figures Page Figure 4-1. CRD Penetration Fracture Toughness Limiting Transients ..... 25 Figure 4-2. Feedwater Nozzle Fracture Toughness Limiting Transient ..... 30 Figure 5-1. Bottom Head P-T Curve for Pressure Test [Curve A] [20°F/hr or less coolant heatup/cooldown] ......................................... 45 Figure 5-2. Upper Vessel P-T Curve for Pressure Test [Curve A] [2 0 °Flhr or less coolant heatup/cooldown] ............................................. 46 Figure 5-3. Beltline P-T Curve for Pressure Test [Curve A] up to 32 EFPY

[2 0 OFlhr or less coolant heatup/cooldown] .......................... 47 Figure 5-4. Beltline P-T Curve for Pressure Test [Curve A] up to 22 EFPY

[2 0 OF/hr or less coolant heatup/cooldown] .......................... 48 Figure 5-5. Bottom Head P-T Curve for Core Not Critical [Curve B]

[IOO0 Flhr or less coolant heatup/cooldown] ........................ 49 Figure 5-6. Upper Vessel P-T Curve for Core Not Critical [Curve B]

[1000F/hr or less coolant heatup/cooldown] ........................ 50 Figure 5-7. Beltline P-T Curve for Core Not Critical [Curve B] up to 32 EFPY [1001F/hr or less coolant heatuplcooldown] ...... 51 Figure 5-8. Beltline P-T Curve for Core Not Critical [Curve B] up to 22 EFPY [IOO0 F/hr or less coolant heatup/cooldown] .......... 52 Figure 5-9. Bottom Head And Upper Vessel Plus Beltline P-T Curves For Core Not Critical [Curve C] Up To 22 EFPY

[100°F/hr or less coolant heatuplcooldown] ........................ 53 Figure 5-10. Bottom Head And Upper Vessel Plus Beltline P-T Curves For Pressure Test [Curve A] Up To 32 EFPY [2 0 °F/hr or less coolant heatup/cooldown] ...................................................... 54 Figure 5-11. Bottom Head And Upper Vessel Plus Beltline P-T Curves For Core Not Critical [Curve B] Up To 32 EFPY [100°F/hr or less coolant heatup/cooldown ..................................................... 55

- vi -

GENE 0000-0000-8763-O1 a Revision 0 Figure 5-12 Bottom Head And Upper Vessel Plus Beltline P-T Curves For Core Critical [Curve C] Up To 32 EFPY [100°F/hr or less coolant heatup/cooldown ..................................................... 56

- vii -

GENE 0000-0000-8763-01a Revision 0 Table of Tables Pa-ge Table 4-1 Calculated Initial RTNDT Values for Perry Vessel Plate Materials ...... 8 Table 4-2 Calculated Initial RTNDT Values for Perry Vessel Nozzle Materials ........................................................................................... 9.

Table 4-3 Calculated Initial RTNDT Values for Perry Vessel Weld and Bolting Materials ........................................................................... 10 Table 4-4a. Perry Beltline ART Values (22 EFPY) ......................................... 13 Table 4-4b. Perry Beltline ART Values (32 EFPY) ......................................... 14 Table 4-5. Summary of the IOCFR50 Appendix G Requirements ............... 17 Table 4-6. Applicable BWR/6 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B ...................................................... 19 Table 4-7. Applicable BWR/6 Discontinuity Components for Use with CRD (Bottom Head) Curves A & B ............................................... 19 Table 5-1. Composite and Individual Curves Used to Construct Composite P-T Curves at 22 and 32 EFPY ................................. 44

- viii -

GENE 0000-0000-8763-Ola Revision 0 1.0 Introduction The pressure-temperature (P-T) curves included in this report have been developed to present steam dome pressure versus minimum vessel metal temperature incorporating appropriate non-beltline limits and irradiation embrittlement effects in the beltline. Complete P-T curves were developed for 22 and 32 effective full power years (EFPY). The P-T curves are provided in Section 5.0 and a tabulation of the curves is included in Appendix B.

The methodology used to generate the P-T curves in this report is presented in Section 4.3 and is similar to the methodology used to generate the P-T curves in November 1996 [1].

Several improvements were made to the P-T curve methodology; the improvements include, but are not limited to the following: 1) The incorporation of ASME Code Case N-640 [4].

2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress [8]. ASME Code Case N-640 allows the use of Kic rather than Kla to determine T-RTNDT. Descriptions of other improvements are included in the P-T curve methodology section. P-T curves are developed using geometry of the RPV shell and discontinuities, the initial reference temperature of nil ductility transition (RTNDT) of the RPV materials, and the adjusted reference temperature (ART) for the beltline materials.

The initial RTNDT is the reference for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code. The Charpy energy data used to determine the initial RTNDT values were tabulated from the Certified Material Test Report (CMTRs). The data and methodology used to determine the initial RTNDT is documented in Section 4.1.

Adjusted Reference Temperature (ART) is the reference temperature including irradiation shift and a margin term. Regulatory Guide 1.99, Rev. 2 [6] provides the methods for calculating ART. The value of ART is a function of RPV 1/4T fluence and beltline material chemistry. The ART calculation, methodology, and ART tables for 22 and 32 EFPY are included in Section 4.2. The 32 EFPY 11/4Tplate fluence value of 4.1x 1018 n/cm 2 used in this report was determined from a new fluence evaluation consistent with the methods defined in Regulatory Guide 1.190 as approved by the NRC [10]. A brief discussion of fluence is GENE 0000-0000-8763-01 a Revision 0 included in Section 4.2.1.2 and a detailed discussion is contained in Appendix H. The chemistry data is discussed in Section 4.2.1.1.

Comprehensive documentation of the RPV discontinuities that are considered in this report is included in Appendix A. This appendix also includes a table to document which non beltline discontinuity curves are used to protect the discontinuities.

Guidelines and requirements for operating and temperature monitoring are included in Appendix C. GE SIL 430, a GE service information letter regarding Reactor Pressure Vessel Temperature Monitoring, is included in Appendix D.

A discussion and the determination of upper shelf energy (USE) is included in Appendix E.

Pressure-temperature curves for a 200°F/hr heatup/cooldown are presented in Appendix F.

Appendix G presents the determination of the beltline region and its impact on fracture toughness. Finally, Appendix H provides a detailed description of the determination of lead factor, peak vessel flux, and fluence.

GENE 0000-0000-8763-01a Revision 0 2.0 Scope of the Analysis This analysis was performed at a reactor power of 3758 MWt (1.05 of 3579 MWt). The methodology used to generate the P-T curves in this report is similar to the methodology used to generate the P-T curves in November 1996 [1]. A detailed description of the P-T curve bases is included in Section 4.3. Several improvements were made to the P-T curve methodology; the improvements include, but are not limited to the following: 1) The incorporation of ASME Code Case N-640 [4]. 2) The use of the Mm calculation in the 1995 ASME Code paragraph G-2214.1 for a postulated defect normal to the direction of maximum stress [8]. ASME Code Case N-640 allows the use of Kc rather than Kla to determine T-RTNDT. Other improvements include, but are not limited to the following:

"* Generation of separate curves for the upper vessel in addition to those generated for the beltline, and bottom head.

"* Comprehensive description of discontinuities used to develop the non-beltline curves (see Appendix A).

The pressure-temperature (P-T) curves are established based on the requirements of 10CFR50, Appendix G [5] to assure that brittle fracture of the reactor vessel is prevented.

Part of the analysis involved in developing the P-T curves is to account for irradiation embrittlement effects in the core region, or beltline. The method used to account for irradiation embrittlement is described in Regulatory Guide 1.99, Rev. 2 [6].

In addition to beltline considerations, there are non-beltline discontinuity limits such as nozzles, penetrations, and flanges that influence the construction of P-T curves. The non beltline limits are based on generic analyses that are adjusted to the maximum reference temperature of nil ductility transition (RTNDT) for the applicable Perry vessel components.

The generic analyses are shown to be bounding for Perry, and the non-beltline limits are discussed in Section 4.3 and are also governed by requirements in [5].

The fluence was evaluated using methods consistent with Reg. Guide 1.190 as approved by the NRC [10]. A brief discussion is included in Section 4.2.1.2 and a detailed discussion is contained in Appendix H. As a result of the fluence evaluation, the upper shelf energy (USE) is evaluated and the results presented in Appendix E.

GENE 0000-0000-8763-01 a Revision 0 Furthermore, curves are included to allow monitoring of the vessel bottom head and upper vessel regions separate from the beltline region. This refinement could minimize heating requirements prior to pressure testing. Operating and temperature monitoring requirements are found in Appendix C. Temperature monitoring requirements and methods are available in GE Services Information Letter (SIL) 430 contained in Appendix D. Pressure Temperature curves for the core not critical and core critical conditions, representing a 200°F/hr heatup/cooldown are presented in Appendix F. Appendix G presents a determination of the beltline region and its impact on fracture toughness. Finally, a detailed discussion of the determination of the lead factor, peak vessel flux, and fluence is presented in Appendix H.

GENE 0000-0000-8763-01 a Revision 0 3.0 Analysis Assumptions The following assumption is made for this analysis:

For end-of-license (32 EFPY) fluence an 80% capacity factor is used to determine the EFPY for a 40-year plant life. The 80% capacity factor is based on the objective to have BWR's available for full power production 80% of the year (Refueling outages, etc. account for 20%

of the year). Perry's capacity factor through power uprate was approximately 68%, and the current cumulative EFPY is approximately 10.3. The next surveillance capsule withdrawal is scheduled for 2012. Even with a capacity factor of 95% until 2012, the cumulative EFPY at the time of withdrawal will be less than 20. Thus, P-T curves for 22 EFPY as presented in this report will be bounding for Perry through the revision of the P-T curves following capsule withdrawal and testing. Therefore, an assumption of an 80% capacity factor is still appropriate.

GENE 0000-0000-8763-01 a Revision 0 4.0 Analysis 4.1 Initial Reference Temperature 4.1.1 Background The initial RTNDT values for all low alloy steel vessel components are needed to develop the vessel P-T limits. The requirements for establishing the vessel component toughness prior to 1972 were per the ASME Code Section III, Subsection NB-2300 as follows:

a. Test specimens shall be longitudinally oriented Charpy V-Notch (CVN) specimens.

b. At the qualification test temperature (specified in the vessel purchase specification), no impact test results shall be less than 25 ft-lb, and the average of three test results shall be at least 30 ft-lb.

c. Pressure tests shall be conducted at a temperature at least 60°F above the qualification test temperature for the vessel materials.

The current requirements used to establish an initial RTNDT value are significantly different.

For plants constructed according to the ASME Code after Summer 1972, the requirements per the ASME Code Section III, Subsection NB-2300 are as follows:

a. Test specimens shall be transversely oriented (normal to the rolling direction)

Charpy V-Notch (CVN) specimens.

b. RTNDT is defined as the higher of the dropweight NDT or 60°F below the temperature at which Charpy V-Notch 50 ft-lb energy and 35 mils lateral expansion are met.

c. Bolt-up in preparation for a pressure test or normal operation shall be performed at or above the highest RTNDT of the materials in the closure flange region or lowest service temperature (LST) of the bolting material, whichever is greater.

GENE 0000-0000-8763-01a Revision 0 4.1.2 Values of Initial RTNDT and Lowest Service Temperature (LST)

To establish the initial RTNDT temperatures for the Perry vessel per the current requirements, calculations were performed in accordance with the ASME Code Section III, Subsection NB-2300.

The first step in calculating RTNDT is to establish the 50 ft-lb transverse test temperature test specimen data (obtained from the CMTRs). For Perry CMTRs, typically three energy values were listed at a given test temperature, corresponding to one set of Charpy tests. The Charpy plate specimens are transversely oriented. All Perry Charpy values were at or above 50 ft-lb and mils lateral expansion (MLE) were at or above the required value. The Charpy impact energy data were taken from the CMTRs [16].

For bolting material, the current ASME Code requirements define the lowest service temperature (LST) as the temperature at which transverse CVN energy of 45 ft-lb and 25 mils lateral expansion (MLE) were achieved. The Charpy impact energy data were taken from the CMTRs [16]. Charpy data for the Perry closure studs did not meet all the 45 ft-lb, 25 MLE requirements at 100 F. Therefore, the LST for the bolting material is at the test temperature + 60OF (70 0 F). The highest RTNDT in the closure flange region is 10°F, for the top head and upper shell materials. Thus, the higher of the LST and the RTNDT + 60°F is 70 0 F, the boltup limit in the closure flange region.

Tables 4-1 thru 4-3 list the calculated initial RTNDT values for the Perry reactor vessel beltline and non-beltline materials. These tabulations include plate, closure flange, nozzle, and weld materials that were considered in generating the P-T curves.

GENE 0000-0000-8763-01 a Revision 0 Table 4-1 Calculated Initial RTInT Values for Perry essel Plate Materials COPNET j ETILJXLT TEST Trans CHARPY ENERGY (Tt-0 DROP R~w. Min Lat S. .. . . ...

(F) Long I I NOT ('F) I (rils)

PLATES &FORGINGS:

Top Head & Flange 48D865-1-1 50 T 96 1 84 1 93 1 -10 1 -20 1 -10 35 Shell Flange Piece #27-1 48D1278-1-1 30 T 53 50 54 -30 -50 -30 35 Head Flange Piece #32-1 48A1187-1-1 30 T 53 50 54 -30 -50 -30 35 B6489-3 50 T 58 61 61 -10 -70 -10 51 Top Head Dome Piece #36-2 B6489-1 70 T 40 Top Head Side Plates 62 154 1 511 10 1 -30 1 10 B6489-2 50 T 57 57 56 -10 -10 -10 50 Piece #36-1 Shell Courses C2451-1 70 T 52 53 60 10 -10 10 47 Upper Shell C2451-2 70 T 58 63 50 10 -30 10 44 Ring #4 Piece #24-1-1/3 41 C3723-2 50 T 53 53 58 -10 -10 -10 C2432-1 s0 T 58 59 58 -10 -30 -10 48 Upper Int. Shell C2432-2 70 T 65 66 59 10 -40 10 57 Ring #3 Piece #23-1-1/3 C2453-1 60 T 77 75 75 0 -10 0 60 C2557-1 70 T 52 50 52 10 -20 10 42 Low-Int. Shell A1155-1 50 T 65 63 67 -10 -20 -10 52 Ring #2 Piece #22-1-1/3 B6270-1 30 T 63 63 64 -30 -40 -30 51 C2448-1 70 T 57 52 50 10 -20 10 46 Lower Shell C2448-2 70 T 54 53 54 10 -20 10 46 Ring #1 Piece #21-1-1/3 A1068-1 70 T 77 56 50 10 -20 10 47 Skirt R0437-1 60 T 52 51 50 0 -10 0 41 Piece #10-2-1/3 B6813-4 50 T 51 51 50 -10 -40 -10 41 Piece #10-1-1 C2128-2A 30 T 50 51 50 -30 -40 -30 38 Piece #10-1-2 Bottom Head Center Plate C2469-1 50 T 54 61 56 -10 -30 -10 50 Piece #13-1 C2469-2 70 T 56 63 68 10 -30 10 51 Side Plate 68 In All114-1 70 7 50 55 10 -10 10 45

1. 13ece . I At 114-1 I 70 T 50 1 68 GENE 0000-0000-8763-Ola Revision 0 Table 4-2 Calculated Initial RTNDT Values for Perry Vessel Nozzle Materials TEST Taso sw DROP R ?T in~ Let COMPONENT HEATIFLUXtLOT TEMP. Trn rCHARPY ENERGY (FT-LB (aeO) WEtG14T

(-) Long (IF) NOT (-) (F) (mils)

Nozzles:

NI Recirc. Outlet Nozzle Piece #49-1-1/2 Q2Q61W 40 T 50 86 98 -20 -20 -20 48 N2 Recirc Inlet Nozzle Piece #52-1-1/10 Q2QLIW 40 T 54 56 59 -20 -20 -20 46 N3 Steam Outlet Nozzle Piece #56-1-1/4 Q2Q61W 40 T 76 93 102 -20 -20 -20 53 N4 Feedwater Nozzle Q2055W 40 T 54 62 50 -20 -20 -20 47 Piece #59-1-1/6 Q2Q61W 40 T 100 97 97 -20 -20 -20 66 N5 Core Spray Nozzle Piece #63-1-1/2 02QL2W 40 T 68 91 81 -20 -20 -20 52 N6 LPCI Nozzle Piece #67-1-1/2 Q2QL3W 40 T 63 71 64 -20 -20 -20 54 Piece #67-1-3 Q2QL3W 40 T 70 75 68 -20 -20 -20 57 N7 Head Sparger Nozzle Piece #71-2 Q2QL8W 40 T 61 56 68 -20 -20 -20 47 N8 Head Spray Nozzle Piece #74-2 Q2QL8W 40 T 61 56 68 -20 -20 -20 47 N9 Jet Pump Instrumentation Nozzle Piece #77-1-1/2 02Q61W 40 T 152 152 90 -20 -20 -20 61 N10 CRD Hyd Sys Rtn Nozzle Piece #80-1 Q2Q51W 40 T 85 61 88 -20 -20 -20 52 Nil Core Diff. Press. & Liquid Control Piece #84-1-1 54318 40 T 85 61 88 -20 -20 -20 52 N15 Drain Nozzle Piece #93-1 719282 30 T 180 209 239 -30 -30 -30 83 N16 Instrumentation Vibration Nozzle Piece #95-1 Q2QL3W 40 T 53 65 50 -20 -20 -20 44 GENE 0000-0000-8763-01 a Revision 0 Table 4-3 Calculated Initial RTNDT Values for Perry Vessel Weld and Bolting Materials TEST TrnorDO NO Mi a COMPON~ENT HEATIFLUXILOT TEMP. C AiRPY E RG T-B TaSOri) RT E t MmGH WELDS:

Vertical Welds Shell Ring #1 BA, BB, BC 5P6214B 10 n/a 56 50 54 -50 -40 -40 41 Shell Ring #2 BD, BE, BF 5P6214B 10 n/a 56 50 54 -50 -40 -40 41 BD. BF 627260 Lot B322A27AE 30 n/a 52 56 51 -30 -40 -30 35 BD. BE, BF 62667? Lot C301A27AF 40 n/a 53 51 54 -20 -40 -20 35 BE 624063 Lot C22BA27A 10 n/a 57 59 68 -50 -60 -50 37 BE 627069 Lot C312A27AG 0 n/a 72 64 78 -60 -60 -60 48 Shell Ring #3 BG, BJ, BK 5P6214B 10 n/a 56 50 54 -50 -40 -40 41 Shell Ring #4 BN, BP, BR 5P6214B 10 n/a 56 50 54 -50 -40 -40 41 Girth Welds Bottom Head to Ring #1: AA- 5P5657 Lot 0342 Ring #1 to Ring #2: AB 4P7216 Lot 0156 (single) 40 n/a 60 72 60 -20 -70 -20 47 4P7216 Lot 0156 (tandem) 40 n/a 80 74 74 -20 -60 -20 59 Ring #2 to Ring #3: AC* 402P3162 Lot H426B27AE 492L4871 Lot A422B27AE 047931 Lot A423B27AG 412L4711 Lot A423027AH 422K8511 Lot G313A27AD 07R458 Lot S403B27AG 0 n/a 59 61 70 -60 -60 -60 51 Ring #3 to Ring #4: AD** 5P5657 Lot 0342 Ring #4 to TODHead: AE- 5P6256 Lot 0342 STUDS: LST 11054 10 n/a 48 49 49 10 OK 83833 10 n/a 46 44 45 70 OK

• This material is not limnting

"-These are non-beltline materials and therefore have no impact on the P-T curves GENE 0000-0000-8763-01a Revision 0 4.2 Adjusted Reference Temperature for Beitline The adjusted reference temperature (ART) of the limiting beitline material is used to adjust the beltline P-T curves to account for irradiation effects. Regulatory Guide 1.99, Revision 2 (Rev 2) provides the methods for determining the ART. The Rev 2 methods for determining the limiting material and adjusting the P-T curves using ART are discussed in this section.

As discussed in Appendix G, the beltline contains the girth weld materials both above and below Shell #2, which represents a slight extension beyond the core region. This is determined by the location on the vessel where the fluence exceeds lx1017 n/cm 2 . An evaluation of ART for the beltline plates and welds were made and summarized in Tables 4-4a and 4-4b for 22 and 32 EFPY, respectively.

4.2.1 Regulatory Guide 1.99, Revision 2 (Rev 2) Methods The value of ART is computed by adding the SHIFT term for a given value of effective full power years (EFPY) to the initial RTNDT. For Rev 2, the SHIFT equation consists of two terms:

SHIFT = ARTNDT + Margin where, ARTNDT = [CF]*f (0.28-0.10 log f) 2 05 Margin = 2(a,2 + 0A ) .

CF = chemistry factor from Tables 1 or 2 of Rev. 2 f = fsurf(e. 24x)/1019 (refer to 4.2.1.2 below)

Margin = 2(0a2 + UA2)0.5 a = standard deviation on initial RTNDT, which is taken to be 0°F (refer to section 4.2.1.1 below).

CYA= standard deviation on ARTNDT, 28°F for welds and 17'F for base material, except that aA need not exceed 0.50 times the ARTNDT value.

ART = Initial RTNDT + SHIFT 4.2.1.1 Chemistry The vessel beltline chemistries were obtained from CMTRs [16] and the USAR [11].

GENE 0000-0000-8763-01 a Revision 0 The copper (Cu) and nickel (Ni) values were used with Tables 1 and 2 of Rev 2, to determine a chemistry factor (CF) per Paragraph 1.1 of Rev 2 for welds and plates, respectively. The margin term, cA, has constant values in Rev 2 of 170 F for plates and 28 0 F for welds. However, 0 A need not be greater than 0.5*ARTNDT. Since the GE method of estimating RTNDT operates on the lowest Charpy energy value to determine the initial RTNOT at the 50 ft-lb level, the value of a, is taken to be 0°F.

4.2.1.2 Fluence The Reactor Pressure Vessel fluence values (fsuf) at 32 EFPY were determined from a new fluence evaluation consistent with methods defined in Reg. Guide 1.190 as approved by the NRC [10] and discussed in detail in Appendix H.

The following 32 EFPY inside surface beltline plate and weld fluences were used:

2 Plate: fsurf = 4.1x 1018 n/cm 2

Weld: f5ur = 4.lxl018 n/cm The 32 EFPY 1/4T fluences are calculated using the following methodology from Rev. 2 [6]

fl/4T = f

• e °24"1"5 Where: 1.5 = % of the minimum beltline shell thickness, 6 inches The resulting 1/4Tfluences are:

18 2 Plate: f1/4T = 2.9x10 n/cm 18 2 Weld: f1I4T = 2.9x10 n/cm 4.2.2 Limiting Beltline Material The limiting beltline material signifies the material that is estimated to receive the greatest embrittlement due to irradiation effects combined with initial RTNDT. Using initial RTNDT, chemistry, and fluence as inputs, Rev 2 was applied to compute ART. Tables 4-4a and 4-4b list values of beltline ART for 22 and 32 EFPY, respectively.

GENE 0000-0000-8763-Ola Revision 0 Table 4-4a. Perry Beltline ART Values (22 EFPY)

Shell #2 Thickness in inches = 6.00 Ratio Peak/ Location = 1.00 32 EFPY Peak I.D. fluence = 4.1E+18 n/cm^2 32 EFPY Peak 114 T fluence = 2.9E+18 n/cm^2 22 EFPY Peak 1/4 T fluence = 2.0E+18 n/cm^2 Shell #2 Vertical Welds Thickness in inches= 6.00 Ratio Peak/ Location = 1.00 32 EFPY Peak I.D. fluence = 4.1E+18 n/cm^2 32 EFPY Peak 1/4 T fluence = 2.9E+18 n/cm^2 22 EFPY Peak 1/4 T fluence = 2.OE+18 n/crn^2 Initial 114 T 22 EFPY 22 EFPY 22 EFPY COMPONENT HEAT OR HEATALOT %Cu %Ni CF RTNoT Fluence A RTNDT 0 GA, Margin Shift ART

- F ncm^2 °F -F 'F PLATES:

Shell #2 Mk 22-1-1 C2557-1 0.060 0.61 37 10 2.OE+18 21 0 10 21 42 52 Mk 22-1-2 B6270-1 0.060 0.63 37 -30 2.OE+18 21 0 10 21 42 12 Mk 22-1-3 Al155-1 0.060 0.63 37 -10 2.OE+18 21 0 10 21 42 32 C2557-1(a) 0.054 0.62 33 10 2.OE+18 19 0 9 19 37 47 WELDS:

Vertical Welds Seam BD, BE, BF 5P6214B 0.020 0.82 27 -40 2.0E+18 15 0 8 15 31 -9 Seam BD, BF 627260 0.060 1.08 82 -30 2.OE+18 46 0 23 46 93 63 Seam BD, BE, BF 626677 0.010 0.85 20 -20 2.0E+18 11 0 6 11 23 3 Seam BE 624063 0.030 1.00 41 -50 2.0E+18 23 0 12 23 46 -4 Seam BE 627069 0.010 0.94 20 -60 2.0E+18 11 0 6 11 23 -37 5P6214BIb) 0025 0.91 34 -40 2n0+18 19 0 10 19 36 -2 (a) Surveillance Plate (Best Estimate Chemistry)

(b) Surveillance Weld (Best Estimate Chemistry)

GENE 0000-0000-8763-O1 a Revision 0 Table 4-4b. Perry Beltline ART Values (32 EFPY)

Shell #2 Thickness in inches = 6.00 Ratio Peak/ Location = 1.00 32 EFPY Peak I.D. fluence = 4.1E+18 n/cm^2 32 EFPY Peak 1/4 T fluence = 2.9E+18 n/cm^2 32 EFPY Peak 1/4 T fluence = 2.9E+18 n/cm^2 Shell #2 Vertical Welds Thickness in inches= 6.00 Ratio Peak/ Location = 1.00 32 EFPY Peak ID. fluence = 4.1E+18 n/cm^2 32 EFPY Peak 1/4 T fluence = 2.9E+18 n/cm^2 32 EFPY Peak 1/4 T fuence= 2.9E+18 n/cm^2 Initial 114 T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF RTNoTr Fluence A RTrmT ci o0 Margin Shift ART

-F n/cm^2 °F -F 'FFF PLATES:

Shell #2 Mk 22-1-1 C2557-1 0.060 0.61 37 10 2.9E+18 24 0 12 24 49 59 Mk 22-1-2 B6270-1 0.060 0.63 37 -30 2.9E+18 24 0 12 24 49 19 Mk 22-1-3 A1155-1 0.060 0.63 37 -10 2.9E+18 24 0 12 24 49 39 C2557-1(a) 0.054 0.62 33 10 2.9E+18 22 0 11 22 43 53 WELDS:

Vertical Welds Seam BD, BE, BF 5P6214B 0.020 0.82 27 -40 2.9E+18 18 0 9 18 36 -4 Seam BD, BF 627260 0.060 1.08 82 -30 2.9E+18 54 0 27 54 108 78 Seam BD, BE, BF 626677 0.010 0.85 20 -20 2.9E+18 13 0 7 13 26 6 Seam BE 624063 0.030 1.00 41 -50 2.9E+18 27 0 13 27 54 4 Seam BE 627069 0.010 0.94 20 -60 2.9E+18 13 0 7 13 26 -34

_ 5P6214B(b) 0.025 0.91 34 -40 2.9E+18 22 0 11 22 45 5 (a) Surveillance Plate (Best Estimate Chemistry)

(b) Surveillance Weld (Best Estimate Chemistry)

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GENE 0000-0000-8763-Ola Revision 0 4.3 Pressure-Temperature Curve Methodology 4.3.1 Background Nuclear Regulatory Commission (NRC) 10CFR50 Appendix G [5] specifies fracture toughness requirements to provide adequate margins of safety during the operating conditions that a pressure-retaining component may be subjected to over its service lifetime.

The ASME Code (Appendix G of Section XI of the ASME Code [8]) forms the basis for the requirements of 10CFR50 Appendix G. The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

There are four vessel regions that should be monitored against the P-T curve operating limits; these regions are defined on the thermal cycle diagram [2]:

"* Closure flange region (Region A)

"* Core beltline region (Region B)

"* Upper vessel (RegionsA & B)

"* Lower vessel (Regions B & C)

The closure flange region includes the bolts, top head flange, and adjacent plates and welds. The core beltline is the vessel location adjacent to the active fuel, such that the 17 neutron fluence is sufficient (>1x10 n/cm2 ) to cause a significant shift of RTNDT. The remaining portion of the vessel (i.e., upper vessel, lower vessel) includes shells, components like the nozzles, and the support skirt; these regions will also be called the non beltline region.

For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr or less for which the curves are applicable. In addition, Appendix F provides the same curves with a coolant heatup and cooldown temperature rate of 2000 F/hr. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. The bounding transients used to develop the curves are described in the sections below. For the hydrostatic pressure and leak test curve, a GENE 0000-0000-8763-01a Revision 0 coolant heatup and cooldown temperature rate of 20°F/hr or less must be maintained at all times.

The P-T curves for the heatup and cooldown operating condition at a given EFPY apply for both the %T and 4"T locations. When combining pressure and thermal stresses, it is usually necessary to evaluate stresses at the 1/4T location (inside surface flaw) and the 3/4T location (outside surface flaw). This is because the thermal gradient tensile stress of interest is in the inner wall during cooldown and is in the outer wall during heatup. However, as a conservative simplification, the thermal gradient stress at the 1T is assumed to be tensile for both heatup and cooldown. This results in the approach of applying the maximum tensile stress at the 1 4T location. This approach is conservative because irradiation effects cause the allowable toughness, Kir, at 1/4Tto be less than that at 33/4T for a given metal temperature.

This approach causes no operational difficulties, since the BWR is at steam saturation conditions during normal operation, well above the heatup/cooldown curve limits.

The applicable temperature is the greater of the 10CFR50 Appendix G minimum temperature requirement and the ASME Appendix G limits. A summary of the requirements is as follows in Table 4-5:

GENE 0000-0000-8763-Ola Revision 0 Table 4-5. Summary of the IOCFR50 Appendix G Requirements Operating Condition and PressureMimuTeprte Requirement

1. Hydrostatic Pressure Test & Leak Test (Core is Not Critical) - Curve A

1. At < 20% of preservice hydrotest Larger of ASME Limits or of pressure highest closure flange region initial RTNDT + 60 0F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of pressure highest closure flange region initial RTNDT + 90°F II. Normal operation (heatup and cooldown),

including anticipated operational occurrences

a. Core not critical - Curve B

1. At < 20% of preservice hydrotest Larger of ASME Limits or of pressure highest closure flange region initial RTNDT + 600 F*

2. At > 20% of preservice hydrotest Larger of ASME Limits or of pressure highest closure flange region initial RTNDT + 120 0 F

b. Core critical - Curve C

1. At < 20% of preservice hydrotest Larger of ASME Limits + 40°F pressure, with the water level or of a.1 within the normal range for power operation

2. At > 20% of preservice hydrotest Larger of ASME Limits + 40°F pressure or of a.2 + 40°F or the minimum permissible temperature for the inservice system hydrostatic pressure test 60'F adder is included by GE as an additional conservative measure as discussed in Section 4.3.2.3 There are four vessel regions that affect the operating limits: the closure flange region, the core beltline region, and the two regions in the remainder of the vessel (i.e., the upper vessel and lower vessel non-beltline regions). The closure flange region limits are controlling at lower pressures primarily because of 10CFR50 Appendix G [5] requirements.

The non-beltline and beltline region operating limits are evaluated according to procedures in 10CFR50 Appendix G [51, ASME Code Appendix G [8], and Welding Research Council (WRC) Bulletin 175 [9]. The beltline region minimum temperature limits are adjusted to account for vessel irradiation.

GENE 0000-0000-8763-01 a Revision 0 GE PROPRIETARY INFORMATION DELETED 4.3.2 P-T Curve Methodology 4.3.2.1 Non-Beitline Regions Non-beltline regions are defined as the vessel locations that are remote from the active fuel and where the neutron fluence is not sufficient (<1.0E17 n/cm 2) to cause any significant shift of RTNDT. Non-beltline components include nozzles, closure flanges, some shell plates, the top and bottom head plates and the control rod drive (CRD) penetrations.

Detailed stress analyses of the non-beltline components were performed for the BWR/6 specifically for the purpose of fracture toughness analysis. The analyses took into account all mechanical loading and anticipated thermal transients. Transients considered include 100 0 F/hr start-up and shutdown, SCRAM, loss of feedwater heaters or flow, loss of recirculation pump flow, and all transients involving emergency core cooling injections.

Primary membrane and bending stresses and secondary membrane and bending stresses due to the most severe of these transients were used according to the ASME Code [8] to develop plots of allowable pressure (P) versus temperature relative to the reference temperature (T - RTNDT). Plots were developed for the limiting BWRI6 components: the feedwater nozzle (FW) and the CRD penetration (bottom head). All other components in the non-beltline regions are categorized under one of these two components as described in Tables 4-6 and 4-7.

GENE 0000-0000-8763-Ola Revision 0 Table 4-6. Applicable BWR/6 Discontinuity Components for Use With FW (Upper Vessel) Curves A & B Discontinuity Identification FW Nozzle LPCI Nozzle CRD HSR Nozzle Core Spray Nozzle Recirculation Inlet Nozzle Steam Outlet Nozzle Support Skirt and Bottom Head Jet Pump Instrumentation Nozzle Shroud Support Attachment (RPV Shell)

Table 4-7. Applicable BWR/6 Discontinuity Components for Use with CRD (Bottom Head) Curves A & B Discontinuity Identification CRD and Bottom Head Vibration Instrumentation Nozzle Core AP and Liquid Control Nozzle (RPV Shell)

Top Head Nozzles Recirculation Outlet Nozzle Main Closure Flange Steam Water Interface Shell Discontinuities The P-T curves for the non-beltline region were conservatively developed for a large BWRJ6 (nominal inside diameter of 251 inches). The analysis is considered appropriate for Perry as the plant specific geometric values are bounded by the generic analysis for a large BWR/6, shown in Sections 4.3.2.1.1 through 4.3.2.1.4. The generic value was adapted to the conditions at Perry by using plant specific RTNDT values for the reactor pressure vessel (RPV). The presence of nozzles and CRD penetration holes of the upper vessel and bottom head, respectively, has made the analysis different from a shell analysis such as the beltline. This was the result of the stress concentrations and higher thermal stress for certain transient conditions experienced by the upper vessel and the bottom head.

GENE 0000-0000-8763-01 a Revision 0 4.3.2.1.1 PressureTest - Non-Beltline, Curve A (Using Bottom Head)

In a finite element analysis [ ], the CRD penetration region was modeled to compute the local stresses for determination of the stress intensity factor, K1. The evaluation was modified to consider the new requirement for Mm as discussed in ASME Code Section XI Appendix G [8] and shown below. The results of that computation were K,= 143.6 ksi-in..2 for an applied pressure of 1593 psig (1563 psig preservice hydrotest pressure at the top of the vessel plus 30 psig hydrostatic pressure at the bottom of the vessel). The computed value of (T - RTNDT) was 84*F.

The limit for the coolant temperature change rate is 20 °F/hr or less.

GENE 0000-0000-8763-Ola Revision 0 The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [8] was based on a thickness of 8.0 inches; hence, t'12 = 2.83. The resulting value obtained was:

Mm 1.85 for ViI<2 M,, = 0.926 ft for 2<- <3.464 = 2.6206 Mm = 3.21 for ft >3.464 Kim is calculated from the equation in Paragraph G-2214.1 [8] and Kib is calculated from the equation in Paragraph G-2214.2 [8]:

Kim = Mm "arpm = ksi-in" 2 Kib = (2/3) Mm, apb = ksi-in1 /2 The total K, is therefore:

KI = 1.5 (Kim+ Kib) + MmB (asr + (2/3)

• CFs) = 143.6 ksi-in/2 This equation includes a safety factor of 1.5 on primary stress. The method to solve for (T -RTNDT) for a specific K, is based on the Kjc equation of Paragraph A-4200 in ASME Appendix A [7]:

(T - RTNDT) = In [(K1 - 33.2) / 20.7341 / 0.02 (T - RTNDT) = In [(144 - 33.2) / 20.7341/ 0.02 (T - RTNDT) = 84°F The generic curve was generated by scaling 143.6 ksi-in'/ 2 by the nominal pressures and calculating the associated (T - RTNDT):

GENE 0000-0000-8763-01 a Revision 0 Pressure Test CRD Penetration K, and (T - RTNDT) as a Function of Pressure Nominal Pressure K, T- RTNDT (psig) (ksi-in"2 ) (OF) 1563 144 84 1400 129 77 1200 111 66 1000 92 52 800 74 33 600 55 3 400 37 -88 The highest RTNDT for the bottom head plates and welds is 100 F, as shown in Tables 4-1 through 4-3.

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GENE 0000-0000-8763-01a Revision 0 Second, the P-T curve is dependent on the calculated K, value, and the K, value is proportional to the stress and the crack depth as shown below:

K, cc o (7ta)" 2 (4-1)

The stress is proportional to R/t and, for the P-T curves, crack depth, a, is t/4. Thus, K, is proportional to R/(t) 12. The generic curve value of R/(t)" 2, based on the generic BWR/6 bottom head dimensions, is:

Generic: R / (t)"2 = 138 / (8)1/2 = 49 inch" 2 (4-2)

The Perry specific bottom head dimensions are R = 130.1875 inches and t= 8 inches minimum [17], resulting in:

Perry specific: R/ (t)'/ 2 = 130.1875/ (8)1/2 = 46.03 inch"12 (4-3)

Since the generic value of R/(t)" 2 is larger, the generic P-T curve is conservative when applied to the Perry bottom head.

4.3.2.1.2 Core Not Critical HeatupiCooldown - Non-Beltilne Curve B (Using Bottom Head)

As discussed previously, the CRD penetration region limits were established primarily for consideration of bottom head discontinuity stresses during pressure testing.

Heatup/cooldown limits were calculated by increasing the safety factor in the pressure testing stresses (Section 4.3.2.1.1) from 1.5 to 2.0.

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GENE 0000-0000-8763-01 a Revision 0 The calculated value of K, for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [8] for comparison with KIR, the material fracture toughness. A safety factor of 2.0 is used for the core not critical condition. Therefore, the K, value for the core not critical condition is (143.6 / 1.5). 2.0 = 191.5 ksi-in" 2.

Therefore, the method to solve for (T - RTNDT) for a specific K, is based on the Kjc equation of Paragraph A-4200 in ASME Appendix A [7] for the core not critical curve:

(T - RTNDT) = In [(KI - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(191.5 - 33.2) / 20.734] / 0.02 (T- RTNDT) = 102°F.

The generic curve was generated by scaling 192 ksi-in 1 2 by the nominal pressures and calculating the associated (T - RTNDT):

Core Not Critical CRD Penetration K, and (T - RTNDT) as a Function of Pressure Nominal Pressure K, T - RTNDT (psig) (ksi-in"2) (OF) 1563 192 102 1400 172 95 1200 147 85 1000 123 73 800 98 57 600 74 33 400 49 -14 GENE 0000-0000-8763-01a Revision 0 Figure 4-1. CRD Penetration Fracture Toughness Limiting Transients GENE 0000-0000-8763-01 a Revision 0 The highest RTNDT for the bottom head plates and welds is 100 F, as shown in Table 4-1.

As discussed in Section 4.3.2.1.1 an evaluation is performed to assure that the CRD discontinuity bounds the other discontinuities that are to be protected by the CRD curve with respect to pressure stresses (see Table 4-6, 4-7, and Appendix A). With respect to thermal stresses, the transients evaluated for the CRD are similar to or more severe than those of the other components being bounded. Therefore, for heatup/cooldown conditions, the CRD penetration provides bounding limits.

4.3.2.1.3 PressureTest - Non-Beltline Curve A (Using Feedwater Nozzle/Upper Vessel Region)

The stress intensity factor, KI, for the feedwater nozzle was computed using the methods from WRC 175 [9] together with the nozzle dimension for a generic 251-inch BWRI6 feedwater nozzle. The result of that computation was K,= 200 ksi-in"2 for an applied pressure of 1563 psig preservice hydrotest pressure.

The respective flaw depth and orientation used in this calculation is perpendicular to the maximum stress (hoop) at a depth of 1/4T through the corner thickness.

To evaluate the results K, is calculated for the upper vessel nominal stress, PR/t, according to the methods in ASME Code Appendix G (Section III or XI). The result is compared to that determined by CBIN in order to quantify the K magnification associated with the stress concentration created by the feedwater nozzles. A calculation of K, is shown below using the BWRI6, 251-inch dimensions:

Vessel Radius, R, 126.7 inches Vessel Thickness, t. 6.1875 inches Vessel Pressure, Pv 1563 psig GENE 0000-0000-8763-01 a Revision 0 Pressure stress: a = PR / t = 1563 psig 126.7

• inches / (6.1875 inches) = 32005 psi. The Dead weight and thermal restraint free end (RFE) stress of 2.967 ksi is conservatively added yielding a = 34.97 ksi. The factor F (a/rn) from Figure A5-1 of WRC-175 is 1.4 where:

a 'A ( tn 2 + tv 2)1/2 =2.36 inches tn= thickness of nozzle = 7.125 inches tv= thickness of vessel = 6.1875 inches rn= apparent radius of nozzle = ri + 0.29 rc=7.09 inches ri= actual inner radius of nozzle = 6.0 inches r= nozzle radius (nozzle corner radius) = 3.75 inches Thus, airm = 2.36 / 7.09 = 0.33. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.33, is 1.4. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 r (lTa)"/2 F(a/r,):

Nominal K, = 1.5 34.97 - (n- 2.36)1l2 . 1.4 = 200 ksi-in112 The method to solve for (T - RTNDT) for a specific K, is based on the Kic equation of Paragraph A-4200 in ASME Appendix A [7] for the pressure test condition:

(T - RTNDT) = In [(KI - 33.2) / 20.734] / 0.02 (T - RTNDT) = In [(200 - 33.2) / 20.734] / 0.02 (T - RTNDT) = 104.2 *F The generic pressure test P-T curve was generated by scaling 200 ksi-in1 /2 by the nominal pressures and calculating the associated (T - RTNDT):

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GENE 0000-0000-8763-01 a Revision 0 The highest RTNDT for the nozzle materials is -20°F as described below. The generic pressure test P-T curve is applied to the Perry feedwater nozzle curve by shifting the P vs.

(T - RTNDT) values above to reflect the RTNDT value of -20 0F.

GENE 0000-0000-8763-Ola Revision 0 Second, the P-T curve is dependent on the K, value calculated. The Perry specific vessel shell [17] and nozzle [18] dimensions applicable to the feedwater nozzle location and K are shown below:

Vessel Radius, R, 120.1875 inches Vessel Thickness, t, 6.0 inches Vessel Pressure, Pv 1563 psig Pressure stress: ar = PR / t = 1563 psig9 120.1875 inches / (6.0 inches) = 31,309 psi. The dead weight and thermal RFE stress of 2.967 ksi is conservatively added yielding a = 34.3 ksi. The factor F (air,) from Figure A5-1 of WRC-1 75 is 1.42 where:

a 1/ ( tn 2+ tv 2)112 = 2.20 inches tn = thickness of nozzle = 6.438 inches tv = thickness of vessel = 6.0 inches rn apparent radius of nozzle = ri + 0.29 rc= 6.94 inches r1 = actual inner radius of nozzle = 6.0 inches r= nozzle radius (nozzle corner radius) = 3.25 inches Thus, afrm = 2.20 / 6.94 = 0.317. The value F(a/rn), taken from Figure A5-1 of WRC Bulletin 175 for an a/ri of 0.317, is 1.42. Including the safety factor of 1.5, the stress intensity factor, K1, is 1.5 a (ra) 1t 2 - F(a/rn):

Nominal K, = 1.5 "34.3. (71 -2.20)1/2 . 1.42 = 192.1 ksi-in1/2 4.3.2.1.4 Core Not Critical Heatup/Cooldown - Non-Beltline Curve B (Using Feedwater Nozzle/Upper Vessel Region)

The feedwater nozzle was selected to represent non-beltline components for fracture toughness analyses because the stress conditions are the most severe experienced in the GENE 0000-0000-8763-01a Revision 0 vessel. In addition to the pressure and piping load stresses resulting from the nozzle discontinuity, the feedwater nozzle region experiences relatively cold feedwater flow in hotter vessel coolant.

Stresses were taken from a finite element analysis done specifically for the purpose of fracture toughness analysis [ ]. Analyses were performed for all feedwater nozzle transients that involved rapid temperature changes. The most severe of these was normal operation with cold 40°F feedwater injection, which is equivalent to hot standby, see Figure 4-2.

Figure 4-2. Feedwater Nozzle Fracture Toughness Limiting Transient The non-beltline curves based on feedwater nozzle limits were calculated according to the methods for nozzles in Appendix 5 of the Welding Research Council (WRC) Bulletin 175 [9].

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GENE 0000-0000-8763-Ola Revision 0 The stress intensity factor for a nozzle flaw under primary stress conditions (Kip) is given in WRC Bulletin 175 Appendix 5 by the expression for a flaw at a hole in a flat plate:

Kip = SF -a (7ta)1/2A F(a/r,) (4-4) where SF is the safety factor applied per WRC Bulletin 175 recommended ranges, and F(a/r,) is the shape correction factor.

Finite element analysis of a nozzle corner flaw was performed to determine appropriate values of F(a/rn) for Equation 4-4. These values are shown in Figure A5-1 of WRC Bulletin 175 [9].

The stresses used in Equation 4-4 were taken from design stress reports for the feedwater nozzle. The stresses considered are primary membrane, ap=, and primary bending, Upb. Secondary membrane, Usm, and secondary bending, csb, stresses are included in the total K, by using ASME Appendix G [8] methods for secondary portion, K1,.

Kis = Mm (usm + (2/3)" GO) (4-5)

In the case where the total stress exceeded yield stress, a plasticity correction factor was applied based on the recommendations of WRC Bulletin 175 Section 5.C.3 [9]. However, the correction was not applied to primary membrane stresses because primary stresses satisfy the laws of equilibrium and are not self-limiting. Kip and Kis are added to obtain the total value of stress intensity factor, K1. A safety factor of 2.0 is applied to primary stresses for core not critical heatup/cooldown conditions.

Once K, was calculated, the following relationship was used to determine (T - RTNDT). The method to solve for (T - RTNDT) for a specific K, is based on the Kic equation of Paragraph A-4200 in ASME Appendix A [7]. The highest RTNDT for the appropriate non-beltline components was then used to establish the P-T curves.

(T - RTNDT) = In [(K, - 33.2) / 20.734] / 0.02 (4-6)

GENE 0000-0000-8763-01 a Revision 0 Example Core Not Critical Heatup/Cooldown Calculation for Feedwater Nozzle/Upper Vessel Region The non-beltline core not critical heatup/cooldown curve was based on the feedwater nozzle analysis, where feedwater injection of 40OF into the vessel while at operating conditions (551.4 0F and 1050 psig) was the limiting normal or upset condition from a brittle fracture perspective. The feedwater nozzle corner stresses were obtained from finite element analysis [ ]. To produce conservative thermal stresses, a vessel and nozzle thickness of 7.5 inches was used in the evaluation. However, a thickness of 7.5 inches is not conservative for the pressure stress evaluation. Therefore, the pressure stress (up,)

was adjusted for the actual vessel thickness of 6.1875 inch (i.e., apm = 20.49 ksi was revised to 20.49 ksi *7.5 inches/6.1875 inches = 24.84 ksi). These stresses, and other inputs used in the generic calculations, are shown below:

GpM = 24.84 ksi GsM = 16.19 ksi -ys = 45.0 ksi t, = 6.1875 inches apb = 0.22 ksi Gsb = 19.04 ksi a = 2.36 inches r, = 7.09 inches t, = 7.125 inches In this case the total stress, 60.29 ksi, exceeds the yield stress, oy, so the correction factor, R, is calculated to consider the nonlinear effects in the plastic region according to the following equation based on the assumptions and recommendation of WRC Bulletin 175 [9].

(The value of specified yield stress is for the material at the temperature under consideration. For conservatism, the temperature assumed for the crack root is the inside surface temperature.)

R = [ays - OpM + ((OtotaI - cry,) / 30)] / (ototai - Gpm) (4-7)

For the stresses given, the ratio, R = 0.583. Therefore, all the stresses are adjusted by the factor 0.583, except for apm. The resulting stresses are:

apm = 24.84 ksi asm = 9.44 ksi Opb = 0.13ksi Gsb =11.10ksi GENE 0000-0000-8763-01a Revision 0 The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [8] was based on the 4a thickness; hence, t1/2 = 3.072. The resulting value obtained was:

Mm = 1.85 for Ft <2 Mm = 0.926 -V" for 2<ft<3.464 = 2.845 Mm = 3.21 for It >3.464 The value F(a/r,), taken from Figure A5-1 of WRC Bulletin 175 for an a/rn of 0.33, therefore, F (a / rn) = 1.4 Kip is calculated from Equation 4-4:

Kip = 2.0 * (24.84 + 0.13) . (T- 2.36)1/2- 1.4 Kip = 190.4 ksi-in" 2 K1s is calculated from Equation 4-5:

Klý = 2.845 * (9.44 + 2/3 - 11.10)

KIs = 47.9 ksi-in 12 The total K, is, therefore, 238.3 ksi-in" 2 .

The total K, is substituted into Equation 4-6 to solve for (T - RTNDT):

(T - RTNDT) = In [(238.3- 33.2) / 20.734] / 0.02 (T- RTNDT) = 115°F The curve was generated by scaling the stresses used to determine the K1; this scaling was performed after the adjustment to stresses above yield. The primary stresses were scaled by the nominal pressures, while the secondary stresses were scaled by the temperature difference of the 40°F water injected into the hot reactor vessel nozzle. In the base case that yielded a K, value of 238 ksi-in 2 , the pressure is 1050 psig and the hot reactor vessel temperature is 551.4 0 F. Since the reactor vessel temperature follows the saturation temperature curve, the secondary stresses are scaled by (Tsaturafo, - 40) / (551.4-40). From K, the associated (T - RTNDT) can be calculated:

GENE 0000-0000-8763-01 a Revision 0 Core Not Critical Feedwater Nozzle K, and (T - RTNDT) as a Function of Pressure Nominal Pressure Saturation Temp. R Kl* (T - RTNDT)

(psig) (OF) (ksi-in"2 ) (OF) 1563 604 0.23 303 128 1400 588 0.34 283 124 1200 557 0.48 257 119 1050 551 0.58 238 115 i1000 546 0.62 232 113 800 520 0.79 206 106 600 489 1.0 181 98 400 448 1.0 138 81

-Note: i-or each change in stress tor each pressure and saturation temperature condition, there is a corresponding change to R that influences the determination of Kl.

The highest non-beltline RTNDT for the feedwater region component at Perry is -20°F as shown in Tables 4-1 through 4-3. The generic curve is applied to the Perry upper vessel by shifting the P vs. (T - RTNDT) values above to reflect the RTNDT value of -200 F.

4.3.2.2 Core Beitline Region The pressure-temperature (P-T) operating limits for the beltline region are determined according to the ASME Code. As the beltline fluence increases with the increase in operating life, the P-T curves shift to a higher temperature.

The stress intensity factors (Ki), calculated for the beltline region according to ASME Code Appendix G procedures [8], were based on a combination of pressure and thermal stresses GENE 0000-0000-8763-Ola Revision 0 for a 1/4T flaw in a flat plate. The pressure stresses were calculated using thin-walled cylinder equations. Thermal stresses were calculated assuming the through-wall temperature distribution of a flat plate; values were calculated for 100°F/hr coolant thermal gradient. The shift value of the most limiting ART material was used to adjust the RTNDT values for the P-T limits. Nozzle N12 (instrumentation nozzle), which is a discontinuity that does not require fracture toughness evaluation (see Appendix A, Table A-2), is located in Shell Ring #2 and is discussed further in Appendix G.

4.3.2.2.1 Beltline Region - PressureTest The methods of ASME Code Section X1, Appendix G [8] are used to calculate the pressure test beltline limits. The vessel shell, with an inside radius (R) to minimum thickness (tmin) ratio of 15, is treated as a thin-walled cylinder. The maximum stress is the hoop stress, given as:

Gm = PR / tmin (4-8)

The stress intensity factor, Kim, is calculated using Paragraph G-2214.1 of the ASME Code.

The calculated value of Kim for pressure test is multiplied by a safety factor (SF) of 1.5, per ASME Appendix G [8] for comparison with K1c, the material fracture toughness. A safety factor of 2.0 is used for the core not critical and core critical conditions.

The relationship between Kic and temperature relative to reference temperature (T - RTNDT) is based on the K1c equation of Paragraph A-4200 in ASME Appendix A [7] for the pressure test condition:

Kim" SF = Kic = 20.734 exp[0.02 (T - RTNDT)] + 33.2 (4-9)

This relationship provides values of pressure versus temperature (from KIR and (T-RTNDT),

respectively).

GE's current practice for the pressure test curve is to add a stress intensity factor, Kit, for a coolant heatup/cooldown rate of 20°F/hr to provide operating flexibility. For the core not critical and core critical condition curves, a stress intensity factor is added for a coolant GENE 0000-0000-8763-01a Revision 0 heatup/cooldown rate of 100°F/hr. The Kit calculation for a coolant heatup/cooldown rate of 100°F/hr is described in Section 4.3.2.2.3 below.

4.3.22.2 Calculationsfor the Beitline Region - PressureTest This sample calculation is for a pressure test pressure of 1100 psig at 32 EFPY. The following inputs were used in the beltline limit calculation:

Adjusted RTNDT = Initial RTNDT + Shift A = -30 + 108 = 780F (Based on ART values in Section 4.2)

Vessel Height H = 852.63 inches Bottom of Active Fuel Height B = 213.5 inches Vessel Radius (to inside of base material) R = 120.1875 inches Minimum Vessel Thickness (without clad) t = 6 inches Pressure is calculated to include hydrostatic pressure for a full vessel:

P = 1100 psi + (H - B) 0.0361 psi/inch = P psig (4-10)

= 1100 + (852.63-213.5)0.0361 = 1123.1 psig Pressure stress:

a = PR/t (4-11)

= 1.123 - 120.1875 / 6.0 = 22.5 ksi The value of Mm for an inside axial postulated surface flaw from Paragraph G-2214.1 [8] was based on a thickness of 6.0 inches (the minimum thickness without cladding); hence, V/2 = 2.449. The resulting value obtained was:

Mm = 1.85 for f: <2 Mm = 0.926 rt for 2< -i_<3.464 = 2.27 Mm = 3.21 for f->3.464 The stress intensity factor for the pressure stress is Kim = Mm- a. The stress intensity factor for the thermal stress, Kit, is calculated as described in Section 4.3.2.2.4 except that the value of "G" is 20°F/hr instead of 100°F/hr.

GENE 0000-0000-8763-Ola Revision 0 Equation 4-9 can be rearranged, and 1.5 Kim substituted for Kic, to solve for (T - RTNDT).

Using the Kic equation of Paragraph A-4200 in ASME Appendix A [7], Kim = 50.95, and Kit= 2.12 for a 20°F/hr coolant heatup/cooldown rate with a vessel thickness, t, that includes cladding:

(T - RTNDT) = ln[(1.5 Kim + Kit - 33.2) / 20.734] / 0.02 (4-12)

= ln[(1.5 50.95 + 2.12 - 33.2) / 20.734] / 0.02

= 39°F T can be calculated by adding the adjusted RTNDT:

T = 39 + 78 = 117'F for P = 1100 psig 4.3.2.2.3 Beltline Region - Core Not CriticalHeatup/Cooldown The beltline curves for core not critical heatup/cooldown conditions are influenced by pressure stresses and thermal stresses, according to the relationship in ASME Section XI Appendix G [8]:

Kic = 2.0

• Kim +Kit (4-13) where Kim is primary membrane K due to pressure and Kit is radial thermal gradient K due to heatup/cooldown.

The pressure stress intensity factor Kim is calculated by the method described above, the only difference being the larger safety factor applied. The thermal gradient stress intensity factor calculation is described below.

The thermal stresses in the vessel wall are caused by a radial thermal gradient that is created by changes in the adjacent reactor coolant temperature in heatup or cooldown conditions. The stress intensity factor is computed by multiplying the coefficient Mt from Figure G-2214-1 of ASME Appendix G [8] by the through-wall temperature gradient ATw, given that the temperature gradient has a through-wall shape similar to that shown in Figure G-2214-2 of ASME Appendix G [8]. The relationship used to compute the through wall ATw is based on one-dimensional heat conduction through an insulated flat plate:

GENE 0000-0000-8763-01 a Revision 0 a 2T(x,t) / a x 2 = 1 / p3(&T(x,t) / at) (4-14) where T(x,t) is temperature of the plate at depth x and time t, and P is the thermal diffusivity.

The maximum stress will occur when the radial thermal gradient reaches a quasi-steady state distribution, so that aT(x,t) / at = dT(t) / dt = G, where G is the coolant heatup/

cooldown rate, normally 100°F/hr. The differential equation is integrated over x for the following boundary conditions:

1. Vessel inside surface (x = 0) temperature is the same as coolant temperature, To.

2. Vessel outside surface (x = C) is perfectly insulated; the thermal gradient dT/dx = 0.

The integrated solution results in the following relationship for wall temperature:

T = Gx 2 / 2P - GCx / 3 + To (4-15)

This equation is normalized to plot (T - To) / AT, versus x / C.

The resulting through-wall gradient compares very closely with Figure G-2214-2 of ASME Appendix G [8]. Therefore, ATw calculated from Equation 4-15 is used with the appropriate Mt of Figure G-2214-1 of ASME Appendix G [8] to compute Kit for heatup and cooldown.

The Mt relationships were derived in the Welding Research Council (WRC) Bulletin 175 [9]

for infinitely long cracks of 1/4T and 1/8T. For the flat plate geometry and radial thermal gradient, orientation of the crack is not important.

4.3.2.2.4 Calculations for the Beltline Region Core Not Critical HeatuplCooldown This sample calculation is for a pressure of 1100 psig for 32 EFPY. The core not critical heatup/cooldown curve at 1100 psig uses the same Kim as the pressure test curve, but with a safety factor of 2.0 instead of 1.5. The increased safety factor is used because the heatup/cooldown cycle represents an operational rather than test condition that necessitates a higher safety factor. In addition, there is a Kit term for the thermal stress. The additional inputs used to calculate Kjt are:

GENE 0000-0000-8763-Ola Revision 0 Coolant heatup/cooldown rate, normally 100°F/hr, G = 100 °F/hr Minimum vessel thickness, including clad thickness, C = 0.511 ft (6.125 inches)

Thermal diffusivity at 550°F (most conservative value), p3 = 0.354 ft 2/ hr [12]

Equation 4-15 can be solved for the through-wall temperature (x = C), resulting in the absolute value of AT for heatup or cooldown of:

AT = GC 2 / 23 (4-16)

= 100- (0.511)2/(2 0.354) = 36.74°F The analyzed case for thermal stress is a %T flaw depth with wall thickness of C. The corresponding value of Mt (=0.287) can be interpolated from ASME Appendix G, Figure G-2214-1 [8]. Thus the thermal stress intensity factor, Kit = Mt

• AT = 10.54, can be calculated. Kim has the same value as that calculated in Section 4.3.2.2.2.

The pressure and thermal stress terms are substituted into Equation 4-9 to solve for (T - RTNDT):

(T - RTNDT) = ln[((2

• Kim + Kit) - 33.2) / 20.734]/0.02 (4-17) ln[(2 - 50.95 + 10.54 - 33.2) / 20.734]/0.02

= 67 OF T can be calculated by adding the adjusted RTNDT:

T =67+78= 145 °F for P = 1100 psig 4.3.2.3 Closure Flange Region 10CFR50 Appendix G [5] sets several minimum requirements for pressure and temperature in addition to those outlined in the ASME Code, based on the closure flange region RTNDT.

In some cases, the results of analysis for other regions exceed these requirements and closure flange limits do not affect the shape of the P-T curves. However, some closure flange requirements do impact the curves, as is true with Perry at low pressures.

The approach used for Perry for the bolt-up temperature was based on a conservative value of (RTNDT + 60), or the LST of the bolting materials, whichever is greater. The 60°F adder is GENE 0000-0000-8763-01a Revision 0 included by GE for two reasons: 1) the pre-1971 requirements of the ASME Code Section III, Subsection A, Paragraph N-220 B included the 60°F adder, and 2) inclusion of the additional 60°F requirement above the RTNDT provides the additional assurance that a flaw size between 0.1 and 0.24 inches is acceptable. As shown in Tables 4-1 through 4-3, the limiting initial RTNDT for the closure flange region is determined from both the Top Head Side Plates and Upper Shell at 100F, and the LST of the closure studs was 70°F; therefore, the bolt-up temperature value used is 70 0 F. This conservatism is appropriate because bolt up is one of the more limiting operating conditions (high stress and low temperature) for brittle fracture. However, temperatures should not be permitted to be lower than 680 F, the fuel shutdown margin, for the reason discussed below.

10CFR50 Appendix G, paragraph IV.A.2 [5] including Table 1, sets minimum temperature requirements for pressure above 20% hydrotest pressure based on the RTNDT of the closure region. Curve A temperature must be no less than (RTNDT + 900 F) and Curve B temperature no less than (RTNDT + 120 0 F).

For pressures below 20% of preservice hydrostatic test pressure (312 psig) and with full bolt preload, the closure flange region metal temperature is required to be at RTNDT or greater as described above. At low pressure, the ASME Code [8] allows the bottom head regions to experience even lower metal temperatures than the flange region RTNDT. However, temperatures should not be permitted to be lower than 68 0 F, the fuel shutdown margin, for the reason discussed below.

The shutdown margin, provided in the Perry Technical Specification, is calculated for a water temperature of 680F. Shutdown margin is the quantity of reactivity needed for a reactor core to reach criticality with the strongest-worth control rod fully withdrawn and all other control rods fully inserted. Although it may be possible to safely allow the water temperature to fall below this 68°F limit, further extensive calculations would be required to justify a lower temperature. The 70°F limit applies when the head is on and tensioned and for the bottom head curve. The 680F limit applies when the head is off, while fuel is in the vessel. When the head is not tensioned and fuel is not in GENE 0000-0000-8763-Ola Revision 0 the vessel, the requirements of IOCFR50 Appendix G [5] do not apply, and there are no limits on the vessel temperatures.

4.3.2.4 Core Critical Operation Requirements of 10CFR50, Appendix G Curve C, the core critical operation curve, is generated from the requirements of 10CFRS0 Appendix G [5], Table 1. Table 1 of [5] requires that core critical P-T limits be 40°F above any Curve A or B limits when pressure exceeds 20% of the pre-service system hydrotest pressure. Curve B is more limiting than Curve A, so limiting Curve C values are at least Curve B plus 40'F for pressures above 312 psig.

Table 1 of 10CFR50 Appendix G [5] indicates that for a BWR with water level within normal range for power operation, the allowed temperature for initial criticality at the closure flange region is (RTNDT + 60 0 F) at pressures below 312 psig. This requirement makes the 10CFR50 Appendix G required minimum criticality temperature 70 0 F, based on an RTNDT of 10°F; however, the minimum criticality temperature is limited to 68 0 F by the fuel shutdown margin, as noted in Section 4.3.2.3. In addition, above 312 psig for Curve C, IOCFR50 Appendix G requires that the temperature must be at least the greater of RTNDT of the closure region + 160°F (10°F+ 160 0 F=170 0 F) or the temperature required for the hydrostatic pressure test (117 0 F for Curve A at 1100 psig). The requirement of closure region RTNDT + 160°F does cause a temperature shift in Curve C at 312 psig.

GENE 0000-0000-8763-01 a Revision 0 5.0 Conclusions and Recommendations The operating limits for pressure and temperature are required for three categories of operation: (a) hydrostatic pressure tests and leak tests, referred to as Curve A; (b) non-nuclear heatup/cooldown and low-level physics tests, referred to as Curve B; and (c) core critical operation, referred to as Curve C.

There are four vessel regions that should be monitored against the P-T curve operating limits, these regions are defined on the thermal cycle diagram [2]:

"* Closure flange region (Region A)

"* Core beltline region (Regions A & B)

"* Upper vessel (Region B)

"* Lower vessel (Regions B & C)

For the core not critical and the core critical curve, the P-T curves specify a coolant heatup and cooldown temperature rate of 100°F/hr or less for which the curves are applicable. P-T curves for the same conditions are also provided with a heatup and cooldown temperature rate of 200°F/hr at 32 EFPY. However, the core not critical and the core critical curves were also developed to bound transients defined on the RPV thermal cycle diagram [2] and the nozzle thermal cycle diagrams [3]. For the hydrostatic pressure and leak test curve, a coolant heatup and cooldown temperature rate of 20°F/hr or less must be maintained at all times.

The P-T curves apply for both heatup/cooldown and for both the 1/T and 3/4T locations because the maximum tensile stress for either heatup or cooldown is applied at the 1/4T location. For beltline curves this approach has added conservatism because irradiation effects cause the allowable toughness, KIr, at 1/4T to be less than that at %T for a given metal temperature.

The following P-T curves were generated for Perry.

Composite P-T curves were generated for each of the Pressure Test and Core Not Critical conditions at 32 effective full power years (EFPY). The composite curves were generated by enveloping the most restrictive P-T limits from the separate GENE 0000-0000-8763-Ola Revision 0 beitline, upper vessel and closure assembly P-T limits. A separate Bottom Head Limits (CRD Nozzle) curve is also individually included with the composite curve for the Pressure Test and Core Not Critical condition.

" Separate P-T curves were developed for the upper vessel, the beltline at 22 and 32 EFPY, and the bottom head for the Pressure Test and Core Not Critical conditions.

" A composite P-T curve was also generated for the Core Critical condition at 22 and 32 EFPY. The composite curves were generated by enveloping the most restrictive P-T limits from the separate beltline, upper vessel, bottom head, and closure assembly P-T limits.

" P-T curves are provided for a heatup/cooldown temperature rate of 200°F/hr. These curves are provided for core not critical and core critical conditions, include a separate bottom head curve, and are contained in Appendix F.

Table 5-1 shows the figure numbers for each P-T curve. A tabulation of the curves is presented in Appendix B.

GENE 0000-0000-8763-01 a Revision 0 Table 5-1. Composite and Individual Curves Used to Construct Composite P-T Curves at 22 and 32 EFPY Figure Numbers Table Numbers for Presentation for Presentation Curve Curve Description of the P-T of the P-T Curves Curves Curve A Bottom Head Limits (CRD Nozzle) Figure 5-1 Table B-1 Upper Vessel Limits (FW Nozzle) Figure 5-2 Table B-1 Beltline Limits for 32 EFPY Figure 5-3 Table B-1 Beltline Limits for 22 EFPY Figure 5-4 Table B-3 Curve B Bottom Head Limits (CRD Nozzle) Figure 5-5 Table B-1 Upper Vessel Limits (FW Nozzle) Figure 5-6 Table B-1 Beltline Limits for 32 EFPY Figure 5-7 Table B-1 Beltline Limits for 22 EFPY Figure 5-8 Table B-3 Curve C Composite Curve for 22 EFPY** Figure 5-9 Table B-3 A, B, & C Composite Curves for 32 EFPY Bottom Head and Composite Figure 5-10 Table B-1 &

Curve A

• Table B-2 Bottom Head and Composite Figure 5-11 Table B-1 &

Curve B

• Table B-2 Composite Curve C ** Figure 5-12 Table B-1 The Composite Curve A & B curve is the more limiting of three limits, 10CFR50 Bolt-up Limits, Upper Vessel Limits (FW Nozzle), and Beltline Limits. A separate Bottom Head Limits (CRD Nozzle) curve is individually included on this figure.

• • The Composite Curve C curve is the more limiting of four limits, 1 OCFR50 Bolt-up Limits, Bottom Head Limits (CRD Nozzle), Upper Vessel Limits (FW Nozzle), and Beltline Limits.

-44 -

GENE 0000-0000-8763-Ola Revision 0 1400 1300 1200 1100

a. 1000 z¸ a.

900 INITIAL RTNDT VALUE IS I

=J 10°F FOR BOTTOM HEAD 800 w

0 I 700 HEATUP/COOLDOWN RATE OF COOLANT

< 20°F/HR 600

,,J 500 Lu 400 a.

300 200 -BOTTOM HEAD LIMITS 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 5-1. Bottom Head P-T Curve for Pressure Test [Curve A]

[20°F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-01a Revision 0 1400 1300 1200 1100 1000 xz a- 900 0 INITIAL RTNDT VALUE IS

-200 F FOR UPPER VESSEL S8 00 Ce O 700 HEATUP/COOLDOWN RATE OF COOLANT S6 z 00 < 20*F/HR S500 S400 LU 300 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 5-2. Upper Vessel P-T Curve for Pressure Test [Curve A]

[20°F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-01a Revision 0 1400 1300 1200 1100 B

g4 1000 IL 900 0

I BELTLINE CURVE w

CU) 800 ADJUSTED AS SHOWN:

u)

EFPY SHIFT (-F) 32 108 O 700 0

I w

HEATUP/COOLDOWN W 600 RATE OF COOLANT Z

<_20°F/HR

500 0 400 300 200 BELTLINE LIMITS 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(*F)

Figure 5-3. Beltline P-T Curve for Pressure Test [Curve A] up to 32 EFPY [20°F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-01 a Revision 0 1400 1300 1200 INITIAL RTNDT VALUE IS

-30°F FOR BELTLINE 1100 S1000 0

I. 900

-J w BELTLINE CURVE 0a 800 ADJUSTED AS SHOWN:

0 w EFPY SHIFT (-F) 22 93 o700 HEATUP/COOLDOWN z RATE OF COOLANT

< 20 °F/HR gn 400 3U 300 200

-BELTLINE LIMITS 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-4. Beltline P-T Curve for Pressure Test [Curve A] up to 22 EFPY [2 0 °F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-Ola Revision 0 1400 1300 1200 1100

• " 1000 IL go S- 900 0 INITIAL RTNDT VALUE IS 10°F FOR BOTTOM HEAD w

) 800

'U o 700 HEATUP/COOLDOWN RATE OF COOLANT

• 600 _<100°F/HR S500

'u 0 400 300 200 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure 5-5. Bottom Head P-T Curve for Core Not Critical [Curve B]

[100°F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-Ola Revision 0 1400 1300 1200 1100 0r 1000 W

0 SINITIAL

-20'F FORRTNDT UPPERVALUE IS VESSEL I

w

' 800 0 700 HEATUP/COOLDOWN RATE OF COOLANT S600 < 100°F/HR

500 w

m 400 w

300 200

-UPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 5-6. Upper Vessel P-T Curve for Core Not Critical [Curve B]

[100 0F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-Ola Revision 0 1400 1300 1200 INITIAL RTNDT VALUE IS

-30°F FOR BELTLINE 1100 S1000 BELTLINE CURVE w

ADJUSTED AS SHOWN:

0S900 EFPY SHIFT (-F) 32 108 800 w

o 700 HEATUP/COOLDOWN RATE OF COOLANT

< 100°F/HR w

0 w

IL 300 200

-BELTLINE LIMITS 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(*F)

Figure 5-7. Beltline P-T Curve for Core Not Critical [Curve B] up to 32 EFPY [100°F/hr or less coolant heatup/cooldown]

-51 -

GENE 0000-0000-8763-01 a Revision 0 1400 1300 1200 1100 CL a.

w 1000 BELTLINE CURVE x ADJUSTED AS SHOWN:

0a. EFPY SHIFT ('F) w 900 I 22 93 a.

I..

800 I-l 0 700 HEATUP/COOLDOWN RATE OF COOLANT 600 _<100°F/HR 500 Ifl 400 U.'

300 200 BELTLINE LIMITS 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 5-8. Beltline P-T Curve for Core Not Critical [Curve B] up to 22 EFPY [lO0 °F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-Ola Revision 0 1400 1300 1200 INITIAL RTNDT VALUES ARE

-30°F FOR BELTLINE,

-207F FOR UPPER VESSEL, 1100 AND 10°F FOR BOTTOM HEAD 1000 w BELTLINE CURVES I

ADJUSTED AS SHOWN:

(L 900 EFPY SHIFT ('F) 0 22 93 S800

'U o 700 HEATUP/COOLDOWN

• 600 RATE OF COOLANT

< 100°F/HR j 500 w

S400 w

a.

300 200 UPPER VESSEL AND BELTLINE LIMITS 100 *.......BOTTOM HEAD CURVE 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 5-9. Bottom Head And Upper Vessel Plus Beltline P-T Curves For Core Critical [Curve C] Up To 22 EFPY

[100°F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-Ola Revision 0 1400 1300 __

1200 _INITIAL RTNOT VALUES ARE

-30°F FOR BELTLINE,

-20'F FOR UPPER VESSEL, 1100 /o AND B 10*F FOR BOTTOM HEAD 10 01000 -BELTLINE CURVES ADJUSTED AS SHOWN:

9. 900 EFPY SHIFT (-F) o 32 108 I

wS 800 -__- _

oo 700 _

-HEATUP/COOLDOWN 6 RATE OF COOLANT 600_-----_<_ 20OF/HR

- 500 BOTTOM UJ HEAD

-n 70OF

• o400 300 UPPER VESSEL 200 AND BELTLINE

7. LIMITS

.. BOTTOM HEAD 100 CURVE 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 5-10. Bottom Head And Upper Vessel Plus Beitline P-T Curves For Pressure Test [Curve A] Up To 32 EFPY 0

[2 °F/hr or less coolant heatup/cooldown]

- 54-

GENE 0000-0000-8763-O1 a Revision 0 1400 1300 1200 INITIAL RTNDT VALUES ARE

-30oF FOR BELTLINE,

-20'F FOR UPPER VESSEL, 1100 AND 10°F FOR BOTTOM HEAD 1000 BELTLINE CURVES ADJUSTED AS SHOWN:

I. 900 EFPY SHIFT (-F) 0 32 108 LJ S800 O 700 HEATUP/COOLDOWN S600 RATE OF COOLANT z <_100°F/HR 3 500 S400 LU 0:

300

-UPPER VESSEL 200 AND BELTLINE LIMITS 100 - ------ BOTTOM HEAD CURVE 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure 5-11. Bottom Head And Upper Vessel Plus Beltline P-T Curves For Core Not Critical [Curve B] Up To 32 EFPY

[100"F/hr or less coolant heatup/cooldown]

GENE 0000-0000-8763-01 a Revision 0 1400 1300 1200 INITIAL RTNDT VALUES ARE

-30*F FOR BELTLINE,

-20*F FOR UPPER VESSEL, 1100 AND IOF FOR BOTTOM HEAD 0

1000 BELTLINE CURVES ADJUSTED AS SHOWN:

=

I.O 900 EFPY SHIFT ('F) o- 32 108 uJ U) 800 U) o 700 HEATUP/COOLDOWN S600 RATE OF COOLANT.

Z < 100°F/HR E.* BFOTTOM S5 00 HE-AD

'U 70*F Un 400 IL a.

300 _

200G UPPER VESSEL 2001 AND BELTLINE LIMITS 100 T ------ BOTTOM HEAD CURVE 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure 5-12. Bottom Head And Upper Vessel Plus Beltline P-T Curves For Core Critical [Curve C] Up To 32 EFPY

[IOO0 Flhr or less coolant heatup/Cooldown]

GENE 0000-0000-8763-01a Revision 0 References

1. L.J. Tilly, "Perry Unit 1 RPV Surveillance Materials Testing and Analysis" GE-NE, San Jose, CA, November 1996, (GE-NE-B1301793-01, Rev 0)

2. GE Drawing Number 762E458, Revision 7, "Reactor Cycles," GE-NED, San Jose, CA.

Perry RPV Thermal Cycle Diagram. (GE Proprietary)

3. GE Drawing Number 166B7307, Revision 6, "Reactor Vessel - Nozzle Thermal Cycles," GE-NED, San Jose, CA. Perry Reactor Vessel Nozzle Thermal Cycle Diagram. (GE Proprietary)

4. "Alternative Reference Fracture Toughness for Development of P-T Limit Curves Section Xl, Division 1," Code Case N-640 of the ASME Boiler & Pressure Vessel Code, Approval Date February 26, 1999.

5. "Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

6. "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

7. "Analysis of Flaws," Appendix A to Section XI of the ASME Boiler & Pressure Vessel Code, 1995 Edition with addenda through 1996.

8. "Fracture Toughness Criteria for Protection Against Failure," Appendix G to Section III or Xl of the ASME Boiler & Pressure Vessel Code,1995 Edition with Addenda through 1996.

9. "PVRC Recommendations on Toughness Requirements for Ferritic Materials,"

Welding Research Council Bulletin 175, August 1972.

10. Safety Evaluation for NEDC-32983P, "General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation", (TAC No. MA9891), NRC Transmittal MFN 01-050 from S. A. Richards to J. F. Klapproth(GE), September 14, 2001

11. Perry Nuclear Power Plant Units 1 & 2 Updated Safety Analysis Report for Unit 1, Section 5.3, Revision 10.

12. "Materials - Properties", Part D to Section II of the ASME Boiler & Pressure Vessel Code, 1995 Edition with addenda through 1996.

13.

14.

GENE 0000-0000-8763-01a Revision 0 15.

16. Reactor Pressure Vessel Certified Material Test Reports for Perry, (GE PO # 205 AE028)

17. CB & I Nuclear Company, Drawing # 1 Revision 8, "Vessel Outline", Chicago, Illinois (GE-NE VPF# 3521-193)

18. CB & I Nuclear Company, Drawing # 59 Revision 3, "N4 Nozzle Forging (Feedwater)",

Chicago, Illinois (GE-NE VPF# 3521-249)

GENE 0000-0000-8763-01 a Revision 0 Appendix A Description of Discontinuities A-1

GENE 0000-0000-8763-01 a Revision 0 Table A-1. Geometric Discontinuities for Perry - BWRI6 - 238" Vessel A-2

GENE 0000-0000-8763-Ola Revision 0 Table A-1. Cont. - Geometric Discontinuities for Perry - BWR/6 - 238" Vessel A-3

GENE 0000-0000-8763-01a Revision 0 Table A-2. - Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Nozzle or Appurtenance nozze Material Reference Remarks Identificaiton Appurtenance Core Differential Nozzles or appurtenances Press & Liquid made from Alloy 600 require Control - Bottom 2, 5, 23, 35. no fracture toughness Nl1 Head SB 166 - Alloy 600 36.48,51 evaluation Instrumentation - Nozzles or appurtenances Penetrations in !2.5" or made from stainless Shell Ring #2 - SA 336 CL F8 - steel require no fracture N12 Elev. 364" Austenitic Steel 2, 7, 24, 47 toughness evaluation Instrumentation - Nozzles or appurtenances Penetrations in *2.5" or made from stainless Shell Ring #3 - SA 336 CL F8 - steel require no fracture N13 Elev. 518" Austenitic Steel 2, 8,24, 46 toughness evaluation Instrumentation - Nozzles or appurtenances Penetrations in *2.5" or made from stainless Shell Ring #4 - SA 336 CL F8 - steel require no fracture N14 Elev. 606" Austenitic Steel 2, 9,24,47 toughness evaluation Nozzles or appurtenances Core Differential made from Alloy 600 require Pressure - Bottom 2, 5, 23, 33. no fracture toughness N18 Head SA 166- Alloy 600 35, 36, 51 evaluation Steam Dryer Nozzles or appurtenances Support Bracket - made from stainless steel Elev. 622 1/4" at 2,9,25,26, require no fracture toughness Mark 108-1 top of bracket SA 182 TP F304 69 evaluation Nozzles or appurtenances made from stainless steel Feedwater Sparger 2, 8, 29, 30, require no fracture toughness Mark 112-1 Bracket TP F304 L 69 evaluation Nozzles or appurtenances made from stainless steel 2,8,31,32, require no fracture toughness Mark 116-1 Core Spray Bracket SA 182 TP F304L 39, 69 evaluation Nozzles or appurtenances

• 2.5" require no fracture Incore Penetrations Not Applicable 34, 42, 44, 57 toughness evaluation Nozzles or appurtenances Control Rod Drive made from Alloy 600 require Penetrations - 40, 42, 44, no fracture toughness 13Bottom Head Alloy 600 50, 51 evaluation A-4

GENE 0000-0000-8763-01a Revision 0 Table A-2. Cont. - Geometric Discontinuities Not Requiring Fracture Toughness Evaluations Nozzle or Nozzle or Appurtenance Appurtenance Reference Remarks Identificaiton Appurtenance Nozzles or appurtenances Shroud Support made from Alloy 600 Ring Material to require no fracture Mark 20 Shell #1 MS 61.0 - SB168 2, 6, 49, 58. 59 toughness evaluation Nozzles or appurtenances Shroud Support made from Alloy 600 Stub Material to require no fracture Mark 17-1 Bottom Head MS 61.0 - SB168 2,58,59,60 toughness evaluation Nozzles or appurtenances

!2.5" or made from Alloy Seal Leak MS 600, SB166, 600 require no fracture N17 Detector Alloy 600 11,61 toughness evaluation Not a pressure boundary component and loads only occur on this component when the reactor is shut down during an outage.

Therefore, no fracture Top Head Lifting toughness evaluation is Mark 43-1 Luqs SA533 GR B CL 1 1,14,62 required Nozzles or appurtenances Refueling !92.5" require no fracture Mark 45 Bellows Bar SA516 GR 70 11,63,64,69 toughness evaluation Nozzles or appurtenances made from stainless steel Guide Rod require no fracture Mark 106-1 Bracket SA182 F304 2,65,66, 69 toughness evaluation Weld buildup is stainless steel; Nozzles or appurtenances therefore, conclude made from stainless steel Surveillance that bracket is require no fracture Bracket Pads stainless steel 2, 7, 67, 69 toughness evaluation Weld buildup is stainless steel; Nozzles or appurtenances therefore, conclude made from stainless steel Jet Pump Riser that bracket is require no fracture Support Pads stainless steel 2, 7, 68, 69 toughness evaluation A-5

GENE 0000-0000-8763-01 a Revision 0 Appendix A References

1. CB&I Nuclear Company, Drawing #1, Revision 8, "Vessel Outline", Chicago, Illinois (GENE VPF #3521-193)

2. CB&I Nuclear Company, Drawing #2, Revision 9, "Vessel Assembly", Chicago, Illinois (GENE VPF #3521-194)

3. CB&I Nuclear Company, Drawing #6, Revision 6, "Seam Details (Shell)", Chicago, Illinois (GENE VPF #3521-198)

4. CB&I Nuclear Company, Drawing #7, Revision 8, "Seam Details (Heads)", Chicago, Illinois (GENE VPF #3521-199)

5. CB&I Nuclear Company, Drawing #11, Revision 7, "Bottom Head Dollar Plates",

Chicago, Illinois (GENE VPF #3521-203)

6. CB&I Nuclear Company, Drawing #21, Revision 4, "#1 Shell Ring Assembly",

Chicago, Illinois (GENE VPF #3521-212)

7. CB&I Nuclear Company, Drawing #22, Revision 6, "#2 Shell Ring Assembly",

Chicago, Illinois (GENE VPF #3521-213)

8. CB&I Nuclear Company, Drawing #23, Revision 6, "#3 Shell Ring Assembly",

Chicago, Illinois (GENE VPF #3521-214)

9. CB&I Nuclear Company, Drawing #24, Revision 5, "#4 Shell Ring Assembly",

Chicago, Illinois (GENE VPF #3521-215)

10. CB&I Nuclear Company, Drawing #28, Revision 7, "Shell Flange Details", Chicago, Illinois (GENE VPF #3521-218)

A-6

GENE 0000-0000-8763-01a Revision 0

11. CB&I Nuclear Company, Drawing #29, Revision 7, "Shell Flange Assembly",

Chicago, Illinois (GENE VPF #3521-219)

12. CB&I Nuclear Company, Drawing #34, Revision 10, "Top Head Flange Assembly",

Chicago, Illinois (GENE VPF #3521-224)

13. CB&I Nuclear Company, Drawing #35, Revision 0, "Top Head Details", Chicago, Illinois (GENE VPF #3521-225)

14. CB&I Nuclear Company, Drawing #36, Revision 2, "Top Head Assembly", Chicago, Illinois (GENE VPF #3521-226)

15. CB&I Nuclear Company, Drawing #111, Revision 3, "Steam Dryer Hold Down Bracket Attachment", Chicago, Illinois (GENE VPF #3521-292)

16. CB&I Nuclear Company, Drawing #165, Revision 1, "As-Built Vessel Seam Locations", Chicago, Illinois (GENE VPF #3521-522)

17. CB&I Nuclear Company, Drawing #166, Revision 1, "Top and Bottom Head As-Built Seam Locations", Chicago, Illinois (GENE VPF #3521-523)

18. CB&I Nuclear Company, Drawing #8, Revision 9, "Seam Details Nozzles and Safe Ends", Chicago, Illinois (GENE VPF #3521-200)

19. CB&I Nuclear Company, Drawing #52, Revision 4, "N2 Nozzle Forging (Recirculation Inlet)", Chicago, Illinois (GENE VPF #3521-242)

20. CB&I Nuclear Company, Drawing #93, Revision 2, "N15 Nozzle Forging (Drain)",

Chicago, Illinois (GENE VPF #3521-278)

21. CB&I Nuclear Company, Drawing #94, Revision 5, "N15 Nozzle Attachment (Drain)",

Chicago, Illinois (GENE VPF #3521-279)

A-7

GENE 0000-0000-8763-01 a Revision 0

22. CB&I Nuclear Company, Drawing #98, Revision 6, "N16 Nozzle Assembly (Vibration)", Chicago, Illinois (GENE VPF #3521-281)

23. CB&I Nuclear Company, Drawing #84, Revision 3 "N11 & N18 Nozzle Forgings (Core Differential Pressure & Liquid Control, & Core Differential Pressure", Chicago, Illinois (GENE VPF #3521-290)

24. CB&I Nuclear Company, Drawing #88, Revision 1, "N12, N13, & N14 Nozzle Forgings (Instrumentation)", Chicago, Illinois (GENE VPF #3521-274)

25. CB&I Nuclear Company, Drawing #108, Revision 1, "Steam Dryer Support Bracket",

Chicago, Illinois (GENE VPF #3521-289)

26. CB&I Nuclear Company, Drawing #109, Revision 4, "Steam Dryer Support Bracket Assembly", Chicago, Illinois (GENE VPF #3521-290)

27. CB&I Nuclear Company, Drawing #110, Revision 3, "Steam Dryer Hold Down Bracket", Chicago, Illinois (GENE VPF #3521-291)

28. CB&I Nuclear Company, Drawing #71, Revision 5, "N7 Nozzle Forging (Head Spare)", Chicago, Illinois (GENE VPF #3521-333)

29. CB&I Nuclear Company, Drawing #112, Revision 1, "Feedwater Sparger Bracket",

Chicago, Illinois (GENE VPF #3521-293)

30. CB&I Nuclear Company, Drawing #113, Revision 4, "Feedwater Sparger Bracket Attachment", Chicago, Illinois (GENE VPF #3521-294)

31. CB&I Nuclear Company, Drawing #116, Revision 1, "Core Spray Bracket", Chicago, Illinois (GENE VPF #3521-297)

32. CB&I Nuclear Company, Drawing #117, Revision 5, "Core Spray Brackets Attachment", Chicago, Illinois (GENE VPF #3521-298)

A-8

GENE 0000-0000-8763-Ola Revision 0

33. CB&I Nuclear Company, Drawing #102, Revision 6, "N18 Nozzle Assembly (Core Differential Pressure)", Chicago, Illinois (GENE VPF #3521-395)

34. CB&I Nuclear Company, Drawing #162, Revision 2, "Incore Plug Rod Assembly",

Chicago, Illinois (GENE VPF #3521-511)

35. CB&I Nuclear Company, Drawing #13, Revision 9, "Bottom Head Assembly",

Chicago, Illinois (GENE VPF #3521-204)

36. CB&I Nuclear Company, Drawing #17, Revision 4, "Differential Pressure & Liquid Control Line Installation", Chicago, Illinois (GENE VPF #5228-041)

37. CB&I Nuclear Company, Drawing #67, Revision 1, "N6 Nozzle Forging (RHR-LPCI Mode)", Chicago, Illinois (GENE VPF #3521-257)

38. CB&I Nuclear Company, Drawing #63, Revision 3, "N5 Nozzle Forging (Core Spray)", Chicago, Illinois (GENE VPF #3521-253)

39. 158B8476, Revision 5, GE-APED, San Jose, California

40. CB&I Nuclear Company, Drawing #169, Revision 0, "CRD Penetration Plan for As Built Data Information", Chicago, Illinois (GENE VPF #3521-525)

41. CB&I Nuclear Company, Drawing #56, Revision 2, "N3 Nozzle Forging (Steam Outlet)", Chicago, Illinois (GENE VPF #3521-246)

42. CB&I Nuclear Company, Drawing #5, Revision 5, "Control Rod Drive & In-core Housing Installation", Chicago, Illinois (GENE VPF #5228-029)

43. CB&I Nuclear Company, Drawing #49, Revision 3, "N1 Recirculation Outlet Nozzle Forging", Chicago, Illinois (GENE VPF #3521-239)

44. CB&I Nuclear Company, Drawing #6, Revision 8, "Control Rod Drive & Incore Housing Details", Chicago, Illinois (GENE VPF #5228-030)

A-9

GENE 0000-0000-8763-01 a Revision 0

45. CB&I Nuclear Company, Drawing #9, Revision 6, "Skirt Knuckle Details", Chicago, Illinois (GENE VPF #3521-201)

46. CB&I Nuclear Company, Drawing #92, Revision 2, "N13 Instrumentation Nozzle Assembly", Chicago, Illinois (GENE VPF #3521-277)

47. CB&I Nuclear Company, Drawing #90, Revision 2, "N12 & N14 Instrumentation Nozzle Assembly", Chicago, Illinois (GENE VPF #3521-275)

48. CB&I Nuclear Company, Drawing #87, Revision 6, "N11 Nozzle Assembly (core differential Pressure & Liquid Control", Chicago, Illinois (GENE VPF #3521-273)

49. CB&I Nuclear Company, Drawing #20, Revision 6, "Shroud Support Assembly",

Chicago, Illinois (GENE VPF #3521-211)

50. CB&I Nuclear Company, Drawing #15, Revision 4, "Control Rod Drive Penetration",

Chicago, Illinois (GENE VPF #3521-206)

51. CB&I Nuclear Company, Drawing #14, Revision 3, "Plan of Bottom Head Penetrations", Chicago, Illinois (GENE VPF #3521-205)

52. CB&I Nuclear Company, Drawing #59, Revision 3, "N4 Nozzle Forging (Feedwater)",

Chicago, Illinois (GENE VPF #3521-249)

53. CB&I Nuclear Company, Drawing #R1, Revision 10, "Weld Seam Identification Records", Chicago, Illinois (GENE VPF #3521-310)

54. CB&I Nuclear Company, Drawing #74, Revision 5, "N8 Nozzle Forging (Head Cooling Spray/RCIC & Vent)", Chicago, Illinois (GENE VPF #3521-261)

55. CB&I Nuclear Company, Drawing #77, Revision 3, "N9 Nozzle Forging (Jet Pump Instrumentation)", Chicago, Illinois (GENE VPF #3521-263)

A-IO

GENE 0000-0000-8763-01 a Revision 0

56. CB&I Nuclear Company, Drawing #80, Revision 1, "N10 Nozzle Forging (CRD HYD System Return)", Chicago, Illinois (GENE VPF #3521-266)

57. CB&I Nuclear Company, Drawing #16, Revision 5, "Incore Penetration Details",

Chicago, Illinois (GENE VPF #3521-207)

58. CB&I Nuclear Company, Drawing #18, Revision 5, "Shroud Support Fabrication Details", Chicago, Illinois (GENE VPF #3521-209)

59. CB&I Nuclear Company, Drawing #19, Revision 6, "Shroud Support Assembly Details", Chicago, Illinois (GENE VPF #3521-210)

60. CB&I Nuclear Company, Drawing #17, Revision 5, "Shroud Support Stubs &

Location", Chicago, Illinois (GENE VPF #3521-208)

61. CB&I Nuclear Company, Drawing #99, Revision 1, "N17 Nozzle Forging (Seal Leak Detector)", Chicago, Illinois (GENE VPF #3521-334)

62. CB&I Nuclear Company, Drawing #43, Revision 2, "Top Head Lifting Lugs", Chicago, Illinois (GENE VPF #3521-233)

63. CB&I Nuclear Company, Drawing #45, Revision 5, "Refueling Bellows Support Assembly", Chicago, Illinois (GENE VPF #3521-235)

64. CB&I Nuclear Company, Drawing #46, Revision 2, "Refueling Bellows Support Details", Chicago, Illinois (GENE VPF #3521-236)

65. CB&I Nuclear Company, Drawing #106, Revision 2, "Guide Rod Bracket", Chicago, Illinois (GENE VPF #3521-287)

66. CB&I Nuclear Company, Drawing #107, Revision 2, "Guide Rod Bracket Attachment", Chicago, Illinois (GENE VPF #3521-288)

A-11

GENE 0000-0000-8763-01 a Revision 0

67. CB&I Nuclear Company, Drawing #115, Revision 5, "Surveillance Bracket Pads",

Chicago, Illinois (GENE VPF #3521-296)

68. CB&I Nuclear Company, Drawing #118, Revision 5, "Jet Pump Riser Support Pads",

Chicago, Illinois (GENE VPF #3521-299)

69. CB&I Nuclear Company, Drawing #131, Revision 5, "Inside Brackets & Refueling Bellows Support As-Built Dimensions", Chicago, Illinois (GENE VPF #3521-385)

70. Design Input Request - Perry P-T Curves, Chuck Wirtz (FirstEnergy), 11/20/01 A-12

GENE 0000-0000-8763-Ola Revision 0 Appendix B Pressure Temperature Curve Data Tabulation B-1

GENE 0000-0000-8763-01 a Revision 0 Table B-I. Perry P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, 5-7, 5-10, 5-11 & 5-12 BOTTOM UPPER BOTTOM UPPER BOTTOM HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE HEAD COMPOSITE CURVE A CURVE A CURVE A CURVEB CURVEB CURVEB CURVEC CURVE C (OFj °F (°F (OF) (OF) (OF) (OF) (OF)

I I I II (OF) (OF) (OF) I 0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 10 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 20 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 30 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 40 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 50 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 60 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 80 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 90 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 100 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 110 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 120 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 130 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 140 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 150 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 160 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 170 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 180 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.0 190 70.0 70.0 70.0 70.0 70.0 70.0 70.0 70.2 200 70.0 70.0 70.0 70.0 70.0 70.0 70.0 72.3 210 70.0 70.0 70.0 70.0 70.0 70.0 70.0 74.3 220 70.0 70.0 70.0 70.0 70.0 70.0 70.0 76.3 230 70.0 70.0 70.0 70.0 70.0 70.0 70.0 78.1 240 70.0 70.0 70.0 70.0 70.0 70.0 70.0 79.9 250 70.0 70.0 70.0 70.0 70.0 70.0 70.0 81.6 260 70.0 70.0 70.0 70.0 70.0 70.0 70.0 83.2 270 70.0 70.0 70.0 70.0 70.0 70.0 70.0 84.8 280 70.0 70.0 70.0 70.0 70.0 70.0 70.0 86.3 290 70.0 70.0 70.0 70.0 70.0 70.0 70.0 87.8 300 70.0 70.0 70.0 70.0 70.0 70.0 70.0 89.2 70.0 70.0 70.0 70.0 70.0 70.0 70.0 90.5 310 312.5 70.0 70.0 70.0 70.0 70.0 70.0 70.0 90.9 312.5 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 320 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 330 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 340 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 350 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 360 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 370 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 380 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 390 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 400 70.0 100.0 70.0 l *0 130 0 70.0 170.0 100.0 1300 400 70.0 1300 B-2

GENE 0000-0000-8763-01a Revision 0 Table B-I. Cont. - Perry P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, 5-7, 5-10, 5-11 & 5-12 BOTTOM UPPER BOTTOM UPPER BOTTOM HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE HEAD COMPOSITE PRESSURE CURVE A CURVE A CURVE A CURVE B CURVE B CURVE B CURVE C CURVE C (PSIG) (F) (OF) (7F) (F) (OF) (OF) (OF) (OF) 410 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 420 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 430 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 440 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 450 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 460 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 470 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 480 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 490 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 500 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 510 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 520 70.0 100.0 100.0 70.0 130.0 130.0 70.0 170.0 530 70.0 100.0 100.0 70.0 130.0 130.0 71.2 170.0 540 70.0 100.0 100.0 70.0 130.0 130.0 73.1 170.0 550 70.0 100.0 100.0 70.0 130.0 130.0 74.9 170.0 560 70.0 100.0 100.0 70.0 130.0 130.0 76.7 170.0 570 70.0 100.0 100.0 70.0 130.0 130.0 78.4 170.0 580 70.0 100.0 100.0 70.0 130.0 130.0 80.0 170.0 590 70.0 100.0 100.0 70.0 130.0 130.0 81.6 170.0 600 70.0 100.0 100.0 70.0 130.0 130.0 83.2 170.0 610 70.0 100.0 100.0 70.0 130.0 130.0 84.7 170.0 620 70.0 100.0 100.0 70.0 130.0 130.0 86.1 170.0 630 70.0 100.0 100.0 70.0 130.0 130.0 87.5 170.0 640 70.0 100.0 100.0 70.0 130.0 130.0 88.9 170.0 650 70.0 100.0 100.0 70.0 130.0 130.0 90.2 170.0 660 70.0 100.0 100.0 70.0 130.0 130.0 91.5 170.0 670 70.0 100.0 100.0 70.0 130.0 130.0 92.8 170.0 680 70.0 100.0 100.0 70.0 130.0 130.0 94.1 170.0 690 70.0 100.0 100.0 70.0 130.0 130.0 95.3 170.0 700 70.0 100.0 100.0 70.0 130.0 130.0 96.4 170.0 710 70.0 100.0 100.0 70.0 130.0 130.0 97.6 170.0 720 70.0 100.0 100.0 70.0 130.0 130.0 98.7 170.0 730 70.0 100.0 100.0 70.0 130.0 130.0 99.8 170.0 740 70.0 100.0 100.0 70.0 130.0 130.0 100.9 170.0 750 70.0 100.0 100.0 70.0 130.0 130.0 102.0 170.0 760 70.0 100.0 100.0 70.0 130.0 130.0 103.0 170.0 770 70.0 100.0 100.0 70.0 130.0 130.0 104.0 170.0 780 70.0 100.0 100.0 70.0 130.0 130.0 105.0 170.0 790 70.0 100.0 100.0 70.0 130.0 130.0 106.0 170.0 800 70.0 100.0 100.0 70.0 130.0 130.0 106.9 170.0 810 70.0 100.0 100.0 70.0 130.0 130.0 107.9 170.0 B-3

GENE 0000-0000-8763-01a Revision 0 Table B-1. Cont. - Perry P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 0F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, 5-7, 5-10, 5-11 & 5-12 BOTTOM UPPER BOTTOM UPPER BOTTOM HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE HEAD COMPOSITE PRESSURE CURVEA CURVE A CURVE A CURVE B CURVE B CURVE B CURVE C CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) (OF) (OF) 820 70.0 100.0 100.0 70.0 130.0 130.0 108.8 170.0 830 70.0 100.0 100.0 70.0 130.0 130.0 109.7 170.0 840 70.0 100.0 100.0 70.6 130.0 130.0 110.6 170.0 850 70.0 100.0 100.0 71.4 130.0 130.0 111.4 170.0 860 70.0 100.0 100.0 72.3 130.0 130.0 112.3 170.0 870 70.0 100.0 100.0 73.1 130.0 130.0 113.1 170.0 880 70.0 100.0 100.0 74.0 130.0 130.6 114.0 170.6 890 70.0 100.0 100.0 74.8 130.0 131.3 114.8 171.3 900 70.0 100.0 100.0 75.6 130.0 132.1 115.6 172.1 910 70.0 100.0 100.3 76.4 130.0 132.8 116.4 172.8 920 70.0 100.0 101.4 77.1 130.0 133.5 117.1 173.5 930 70.0 100.0 102.4 77.9 130.0 134.2 117.9 174.2 940 70.0 100.0 103.4 78.7 130.0 135.0 118.7 175.0 950 70.0 100.0 104.4 79.4 130.0 135.6 119.4 175.6 960 70.0 100.0 105.3 80.1 130.0 136.3 120.1 176.3 970 70.0 100.0 106.3 80.9 130.0 137.0 120.9 177.0 980 70.0 100.0 107.2 81.6 130.0 137.7 121.6 177.7 990 70.0 100.0 108.1 82.3 130.0 138.3 122.3 178.3 1000 70.0 100.0 109.0 83.0 130.0 139.0 123.0 179.0 1010 70.0 100.0 109.9 83.6 130.0 139.6 123.6 179.6 1020 70.0 100.0 110.7 84.3 130.0 140.3 124.3 180.3 1030 70.0 100.0 111.6 85.0 130.0 140.9 125.0 180.9 1040 70.0 100.0 112.4 85.6 130.0 141.5 125.6 181.5 1050 70.0 100.0 113.2 86.3 130.0 142.1 126.3 182.1 1060 70.0 100.0 114.0 86.9 130.0 142.7 126.9 182.7 1070 70.0 100.0 114.8 87.5 130.0 143.3 127.5 183.3 1080 70.0 100.0 115.6 88.2 130.0 143.9 128.2 183.9 1090 70.0 100.0 116.4 88.8 130.0 144.5 128.8 184.5 1100 70.0 100.0 117.1 89.4 130.0 145.1 129.4 185.1 1105 70.0 100.0 117.5 89.7 130.0 145.3 129.7 185.3 1110 70.0 100.0 117.9 90.0 130.0 145.6 130.0 185.6 1120 70.6 100.0 118.6 90.6 130.0 146.2 130.6 186.2 1130 71.2 100.0 119.3 91.2 130.0 146.8 131.2 186.8 1140 71.9 100.0 120.0 91.7 130.0 147.3 131.7 187.3 1150 72.5 100.0 120.7 92.3 130.0 147.8 132.3 187.8 1160 73.1 100.0 121.4 92.9 130.0 148.4 132.9 188.4 1170 73.8 100.0 122.1 93.4 130.0 148.9 133.4 188.9 1180 74.4 100.0 122.8 94.0 130.0 149.4 134.0 189.4 1190 75.0 100.0 123.5 94.5 130.0 150.0 134.5 190.0 1200 75.6 100.0 124.1 95.1 130.0 150.5 135.1 190.5 B-4

GENE 0000-0000-8763-01a Revision 0 Table B-1. Cont. - Perry P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-1, 5-2, 5-3, 5-5, 5-6, 5-7, 5-10, 5-11 & 5-12 BOTTOM UPPER BOTTOM UPPER BOTTOM HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE HEAD COMPOSITE PRESSURE CURVEA CURVEA CURVEA CURVEB CURVEB CURVEB CURVEC CURVEC (PSIG) (*F) (°F) (oF) °F) ('F) ('F) (oF) (*F) 1210 76.2 100.0 124.8 95.6 130.0 151.0 135.6 191.0 1220 76.8 100.0 125.4 96.2 130.0 151.5 136.2 191.5 1230 77.3 100.0 126.0 96.7 130.0 152.0 136.7 192.0 1240 77.9 100.0 126.7 97.2 130.0 152.5 137.2 192.5 1250 78.5 100.0 127.3 97.7 130.0 153.0 137.7 193.0 1260 79.0 100.0 127.9 98.2 130.0 153.5 138.2 193.5 1270 79.6 100.0 128.5 98.7 130.0 154.0 138.7 194.0 1280 80.1 100.0 129.1 99.2 130.0 154.4 139.2 194.4 1290 80.7 100.0 129.7 99.7 130.0 154.9 139.7 194.9 1300 81.2 100.0 130.2 100.2 130.0 155.4 140.2 195.4 1310 81.7 100.0 130.8 100.7 130.0 155.8 140.7 195.8 1320 82.3 100.0 131.4 101.2 130.0 156.3 141.2 196.3 1330 82.8 100.0 131.9 101.6 130.0 156.7 141.6 196.7 1340 83.3 100.0 132.5 102.1 130.0 157.2 142.1 197.2 1350 83.8 100.0 133.1 102.6 130.0 157.6 142.6 197.6 1360 84.3 100.0 133.6 103.0 130.0 158.1 143.0 198.1 1370 84.8 100.0 134.1 103.5 130.0 158.5 143.5 198.5 1380 85.3 100.0 134.7 103.9 130.0 159.0 143.9 199.0 1390 85.8 100.0 135.2 104.4 130.0 159.4 144.4 199.4 1400 86.3 100.0 135.7 104.8 130.0 159.8 144.8 199.8 B-5

GENE 0000-0000-8763-01 a Revision 0 Table B2. Perry Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-10 and 5-11 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

PRESSURE HEAD BELTLINE AT 32 HEAD BELTLINE AT 32 (PSIG) CURVE A EFPY CURVE A CURVEB EFPY CURVE B

( 0 F) (OF) (0 F) ( 0 F) 0 70.0 70.0 70.0 70.0 10 70.0 70.0 70.0 70.0 20 70.0 70.0 70.0 70.0 30 70.0 70.0 70.0 70.0 40 70.0 70.0 70.0 70.0 50 70.0 70.0 70.0 70.0 60 70.0 70.0 70.0 70.0 70 70.0 70.0 70.0 70.0 80 70.0 70.0 70.0 70.0 90 70.0 70.0 70.0 70.0 100 70.0 70.0 70.0 70.0 110 70.0 70.0 70.0 70.0 120 70.0 70.0 70.0 70.0 130 70.0 70.0 70.0 70.0 140 70.0 70.0 70.0 70.0 150 70.0 70.0 70.0 70.0 160 70.0 70.0 70.0 70.0 170 70.0 70.0 70.0 70.0 180 70.0 70.0 70.0 70.0 190 70.0 70.0 70.0 70.0 200 70.0 70.0 70.0 70.0 210 70.0 70.0 70.0 70.0 220 70.0 70.0 70.0 70.0 230 70.0 70.0 70.0 70.0 240 70.0 70.0 70.0 70.0 250 70.0 70.0 70.0 70.0 260 70.0 70.0 70.0 70.0 270 70.0 70.0 70.0 70.0 280 70.0 70.0 70.0 70.0 290 70.0 70.0 70.0 70.0 300 70.0 70.0 70.0 70.0 310 70.0 70.0 70.0 70.0 312.5 70.0 70.0 70.0 70.0 312.5 70.0 100.0 70.0 130.0 320 70.0 100.0 70.0 130.0 330 70.0 100.0 70.0 130.0 340 70.0 100.0 70.0 130.0 350 70.0 100.0 70.0 130.0 B-6

GENE 0000-0000-8763-01a Revision 0 Table B2. Cont. - Perry Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B &C and 20 °F/hr for Curve A FOR FIGURES 5-10 and 5-11 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

PRESSURE HEAD BELTLINE AT 32 HEAD BELTLINE AT 32 (PSIG) CURVE A EFPY CURVE A CURVE B EFPY CURVE B (OF) (OF) (OF) (OF) 360 70.0 100.0 70.0 130.0 370 70.0 100.0 70.0 130.0 380 70.0 100.0 70.0 130.0 390 70.0 100.0 70.0 130.0 400 70.0 100.0 70.0 130.0 410 70.0 100.0 70.0 130.0 420 70.0 100.0 70.0 130.0 430 70.0 100.0 70.0 130.0 440 70.0 100.0 70.0 130.0 450 70.0 100.0 70.0 130.0 460 70.0 100.0 70.0 130.0 470 70.0 100.0 70.0 130.0 480 70.0 100.0 70.0 130.0 490 70.0 100.0 70.0 130.0 500 70.0 100.0 70.0 130.0 510 70.0 100.0 70.0 130.0 520 70.0 100.0 70.0 130.0 530 70.0 100.0 70.0 130.0 540 70.0 100.0 70.0 130.0 550 70.0 100.0 70.0 130.0 560 70.0 100.0 70.0 130.0 570 70.0 100.0 70.0 130.0 580 70.0 100.0 70.0 130.0 590 70.0 100.0 70.0 130.0 600 70.0 100.0 70.0 130.0 610 70.0 100.0 70.0 130.0 620 70.0 100.0 70.0 130.0 630 70.0 100.0 70.0 130.0 640 70.0 100.0 70.0 130.0 650 70.0 100.0 70.0 130.0 660 70.0 100.0 70.0 130.0 670 70.0 100.0 70.0 130.0 680 70.0 100.0 70.0 130.0 690 70.0 100.0 70.0 130.0 700 70.0 100.0 70.0 130.0 710 70.0 100.0 70.0 130.0 720 70.0 100.0 70.0 130.0 730 70.0 1i0n0 70.0 130.0 B-7

GENE 0000-0000-8763-01 a Revision 0 Table B2. Cont. - Perry Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 0F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-10 and 5-11 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

PRESSURE HEAD BELTLINE AT 32 HEA.D BELTLINE AT 32 EFPY CURVE A CURVE B EFPY CURVE B (PSIG) CURVE A (OF) (OF) ( 0 F) (0F) 740 70.0 100.0 70.0 130.0 750 70.0 100.0 70.0 130.0 760 70.0 100.0 70.0 130.0 770 70.0 100.0 70.0 130.0 780 70.0 100.0 70.0 130.0 790 70.0 100.0 70.0 130.0 800 70.0 100.0 70.0 130.0 810 70.0 100.0 70.0 130.0 820 70.0 100.0 70.0 130.0 830 70.0 100.0 70.0 130.0 840 70.0 100.0 70.6 130.0 850 70.0 100.0 71.4 130.0 860 70.0 100.0 72.3 130.0 870 70.0 100.0 73.1 130.0 880 70.0 100.0 74.0 130.6 890 70.0 100.0 74.8 131.3 900 70.0 100.0 75.6 132.1 910 70.0 100.3 76.4 132.8 920 70.0 101.4 77.1 133.5 930 70.0 102.4 77.9 134.2 940 70.0 103.4 78.7 135.0 950 70.0 104.4 79.4 135.6 960 70.0 105.3 80.1 136.3 970 70.0 106.3 80.9 137.0 980 70.0 107.2 81.6 137.7 990 70.0 108.1 82.3 138.3 1000 70.0 109.0 83.0 139.0 1010 70.0 109.9 83.6 139.6 1020 70.0 110.7 84.3 140.3 1030 70.0 111.6 85.0 140.9 1040 70.0 112.4 85.6 141.5 1050 70.0 113.2 86.3 142.1 1060 70.0 114.0 86.9 142.7 1070 70.0 114.8 87.5 143.3 1080 70.0 115.6 88.2 143.9 1090 70.0 116.4 88.8 144.5 1100 70.0 117.1 89.4 145.1 1105 70.0 89.7 145.3 117.5 1105 70.0 145 3 B-8

GENE 0000-0000-8763-01a Revision 0 Table B2. Cont. - Perry Composite P-T Curve Values for 32 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 0F/hr for Curve A FOR FIGURES 5-10 and 5-11 BOTTOM UPPER RPV & BOTTOM UPPER RPV &

PRESSURE HEAD BELTLINE AT 32 HEAD BELTLINE AT 32 (PSIG) CURVE A EFPY CURVE A CURVE B EFPY CURVE B (0 F) (0 F) (0 F) (0 F) 1110 70.0 117.9 90.0 145.6 1120 70.6 118.6 90.6 146.2 1130 71.2 119.3 91.2 146.8 1140 71.9 120.0 91.7 147.3 1150 72.5 120.7 92.3 147.8 1160 73.1 121.4 92.9 148.4 1170 73.8 122.1 93.4 148.9 1180 74.4 122.8 94.0 149.4 1190 75.0 123.5 94.5 150.0 1200 75.6 124.1 95.1 150.5 1210 76.2 124.8 95.6 151.0 1220 76.8 125.4 96.2 151.5 1230 77.3 126.0 96.7 152.0 1240 77.9 126.7 97.2 152.5 1250 78.5 127.3 97.7 153.0 1260 79.0 127.9 98.2 153.5 1270 79.6 128.5 98.7 154.0 1280 80.1 129.1 99.2 154.4 1290 80.7 129.7 99.7 154.9 1300 81.2 130.2 100.2 155.4 1310 81.7 130.8 100.7 155.8 1320 82.3 131.4 101.2 156.3 1330 82.8 131.9 101.6 156.7 1340 83.3 132.5 102.1 157.2 1350 83.8 133.1 102.6 157.6 1360 84.3 133.6 103.0 158.1 1370 84.8 134.1 103.5 158.5 1380 85.3 134.7 103.9 159.0 1390 85.8 135.2 104.4 159.4 1400 86.3 135.7 104.8 159.8 B-9

GENE 0000-0000-8763-01 a Revision 0 TABLE B-3. Perry P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-4, 5-8 and 5-9 BOTTOM BELTLINE BELTLINE HEAD COMPOSITE PRESSURE CURVE A CURVE B CURVE C CURVE C (PSIG) (OF) ( 0 F) (OF) (°F) 0 70.0 70.0 70.0 70.0 10 70.0 70.0 70.0 70.0 20 70.0 70.0 70.0 70.0 30 70.0 70.0 70.0 70.0 40 70.0 70.0 70.0 70.0 50 70.0 70.0 70.0 70.0 60 70.0 70.0 70.0 70.0 70 70.0 70.0 70.0 70.0 80 70.0 70.0 70.0 70.0 90 70.0 70.0 70.0 70.0 100 70.0 70.0 70.0 70.0 110 70.0 70.0 70.0 70.0 120 70.0 70.0 70.0 70.0 130 70.0 70.0 70.0 70.0 140 70.0 70.0 70.0 70.0 150 70.0 70.0 70.0 70.0 160 70.0 70.0 70.0 70.0 170 70.0 70.0 70.0 70.0 180 70.0 70.0 70.0 70.0 190 70.0 70.0 70.0 70.2 200 70.0 70.0 70.0 72.3 210 70.0 70.0 70.0 74.3 220 70.0 70.0 70.0 76.3 230 70.0 70.0 70.0 78.1 240 70.0 70.0 70.0 79.9 250 70.0 70.0 70.0 81.6 260 70.0 70.0 70.0 83.2 270 70.0 70.0 70.0 84.8 280 70.0 70.0 70.0 86.3 290 70.0 70.0 70.0 87.8 300 70.0 70.0 70.0 89.2 310 70.0 70.0 70.0 90.5 312.5 70.0 70.0 70.0 90.9 312.5 100.0 130.0 70.0 170.0 320 100.0 130.0 70.0 170.0 330 100.0 130.0 70.0 170.0 340 100.0 130.0 70.0 170.0 350 100.0 130.0 70.0 170.0 B-10

GENE 0000-0000-8763-01a Revision 0 TABLE B-3. Perry P-T Curve Values for 22 EFPY Required Coolant Temperatures at 200 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-4, 5-8 and 5-9 B-11

GENE 0000-0000-8763-01 a Revision 0 TABLE B-3. Cont - Perry P-T Curve Values for 22 EFPY Required Coolant Temperatures at 100 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-4, 5-8 and 5-9 BOTTOM BELTLINE BELTLINE HEAD COMPOSITE PRESSURE CURVE A CURVE B CURVE C CURVE C (PSIG) ( 0 F) ( 0 F) (OF) (°F) 740 100.0 130.0 100.9 170.0 750 100.0 130.0 102.0 170.0 760 100.0 130.0 103.0 170.0 770 100.0 130.0 104.0 170.0 780 100.0 130.0 105.0 170.0 790 100.0 130.0 106.0 170.0 800 100.0 130.0 106.9 170.0 810 100.0 130.0 107.9 170.0 820 100.0 130.0 108.8 170.0 830 100.0 130.0 109.7 170.0 840 100.0 130.0 110.6 170.0 850 100.0 130.0 111.4 170.0 860 100.0 130.0 112.3 170.0 870 100.0 130.0 113.1 170.0 880 100.0 130.0 114.0 170.0 890 100.0 130.0 114.8 170.0 900 100.0 130.0 115.6 170.0 910 100.0 130.0 116.4 170.0 920 100.0 130.0 117.1 170.0 930 100.0 130.0 117.9 170.0 940 100.0 130.0 118.7 170.0 950 100.0 130.0 119.4 170.0 960 100.0 130.0 120.1 170.0 970 100.0 130.0 120.9 170.0 980 100.0 130.0 121.6 170.0 990 100.0 130.0 122.3 170.0 1000 100.0 130.0 123.0 170.0 1010 100.0 130.0 123.6 170.0 1020 100.0 130.0 124.3 170.0 1030 100.0 130.0 125.0 170.0 1040 100.0 130.0 125.6 170.0 1050 100.0 130.0 126.3 170.0 1060 100.0 130.0 126.9 170.0 1070 100.0 130.0 127.5 170.0 1080 100.6 130.0 128.2 170.0 1090 101.4 130.0 128.8 170.0 1100 102.1 130.1 129.4 170.1 1105 102.5 130.3 129.7 170.3 1110 102.9 130.6 130.0 170.6 1120 103.6 131.2 130.6 171.2 1130 104.3 131.8 131.2 171.8 B-12

GENE 0000-0000-8763-Ola Revision 0 TABLE B-3. Perry P-T Curve Values for 22 EFPY Required Coolant Temperatures at 200 °F/hr for Curves B & C and 20 °F/hr for Curve A FOR FIGURES 5-4, 5-8 and 5-9 BOTTOM BELTLINE BELTLINE HEAD COMPOSITE PRESSURE CURVE A CURVE B CURVE C CURVE C (PSIG) (OF) (OF) (OF) (OF) 1140 105.0 132.3 131.7 172.3 1150 105.7 132.8 132.3 172.8 1160 106.4 133.4 132.9 173.4 1170 107.1 133.9 133.4 173.9 1180 107.8 134.4 134.0 174.4 1190 108.5 135.0 134.5 175.0 1200 109.1 135.5 135.1 175.5 1210 109.8 136.0 135.6 176.0 1220 110.4 136.5 136.2 176.5 1230 111.0 137.0 136.7 177.0 1240 111.7 137.5 137.2 177.5 1250 112.3 138.0 137.7 178.0 1260 112.9 138.5 138.2 178.5 1270 113.5 139.0 138.7 179.0 1280 114.1 139.4 139.2 179.4 1290 114.7 139.9 139.7 179.9 1300 115.2 140.4 140.2 180.4 1310 115.8 140.8 140.7 180.8 1320 116.4 141.3 141.2 181.3 1330 116.9 141.7 141.6 181.7 1340 117.5 142.2 142.1 182.2 1350 118.1 142.6 142.6 182.6 1360 118.6 143.1 143.0 183.1 1370 119.1 143.5 143.5 183.5 1380 119.7 144.0 143.9 184.0 1390 120.2 144.4 144.4 184.4 1400 120.7 144.8 144.8 184.8 B-13

GENE 0000-0000-8763-01 a Revision 0 Appendix C Operating and Temperature Monitoring Requirements C-1

GENE 0000-0000-8763-Ola Revision 0 C.1 Non-Beltline Monitoring During Pressure Tests It is likely that, during leak and hydrostatic pressure testing, the bottom head temperature may be significantly cooler than the beltline. This condition can occur in the bottom head when the recirculation pumps are operating at low speed, or are off, and injection through the control rod drives is used to pressurize the vessel. By using a bottom head curve, the required test temperature at the bottom head could be lower than the required test temperature at the beltline, avoiding the necessity of heating the bottom head to the same requirements of the vessel beltline.

One condition on monitoring the bottom head separately is that it must be demonstrated that the vessel beltline temperature can be accurately monitored during pressure testing. An experiment has been conducted at a BWR-4 that showed that thermocouples on the vessel near the feedwater nozzles, or temperature measurements of water in the recirculation loops provide good estimates of the beltline temperature during pressure testing. Thermocouples on the RPV flange to shell junction outside surface should be used to monitor compliance with upper vessel curves.

Thermocouples on the bottom head outside surface should be used to monitor compliance with bottom head curves. A description of these measurements is given in GE SIL 430, attached in Appendix D. First, however, it should be determined whether there are significant temperature differences between the beltline region and the bottom head region.

C.2 Determining Which Curve to Follow The following subsections outline the criteria needed for determining which curve is governing during different situations. The application of the P-T curves and some of the assumptions inherent in the curves to plant operation is dependent on the proper monitoring of vessel temperatures. A discussion of monitoring of vessel temperatures can be found Appendix D.

C-2

GENE 0000-0000-8763-01 a Revision 0 C.2.1 Curve A: Pressure Test Curve A should be used during pressure tests at times when the coolant temperature is changing by <20 0 F per hour. If the coolant is experiencing a higher heating or cooling rate in preparation for or following a pressure test, Curve B applies.

C.2.2 Curve B: Non-Nuclear HeatuplCooldown Curve B should be used whenever Curve A or Curve C do not apply. In other words, the operator must follow this curve during times when the coolant is heating or cooling faster than 20°F per hour during a hydrotest and when the core is not critical.

C.2.3 Curve C: Core Critical Operation The operator must comply with this curve whenever the core is critical. An exception to this principle is for low-level physics tests; Curve B must be followed during these situations.

C.3 Reactor Operation Versus Operating Limits For most reactor operating conditions, coolant pressure and temperature are at saturation conditions, which are well into the acceptable operating area (to the right of the P-T curves). The operations where P-T curve compliance is typically monitored closely are planned events, such as vessel boltup, leakage testing and startup/shutdown operations, where operator actions can directly influence vessel pressures and temperatures.

The most severe unplanned transients relative to the P-T curves are those which result from SCRAMs, which sometimes include recirculation pump trips. Depending on operator responses following pump trip, there can be cases where stratification of colder water in the bottom head occurs while the vessel pressure is still relatively high. Experience with such events has shown that operator action is necessary to avoid P-T curve exceedance, but there is adequate time for operators to respond.

In summary, there are several operating conditions where careful monitoring of P-T conditions against the curves is needed:

"* Head flange boltup

"* Leakage test (Curve A compliance)

C-3

GENE 0000-0000-8763-01 a Revision 0

"* Startup (coolant temperature change of less than or equal to 100°F in one hour period heatup)

"* Shutdown (coolant temperature change of less than or equal to 100OF in one hour period cooldown)

" Recirculation pump trip, bottom head stratification (Curve B compliance)

C-4

GENE 0000-0000-8763-01 a Revision 0 Appendix D GE SIL 430 D-1

GENE 0000-0000-8763-Ola Revision 0 September 27, 1985 SIL No. 430 REACTOR PRESSURE VESSEL TEMPERATURE MONITORING Recently, several BWR owners with plants in initial startup have had questions concerning primary and alternate reactor pressure vessel (RPV) temperature monitoring measurements for complying with RPV brittle fracture and thermal stiress requirements. As such, the purpose of this Service Information Letter is to provide a summary of RPV temperature monitoring measurements, their primary and alternate uses and their limitations (See the attached table). Of basic concern is temperature monitoring to comply with brittle fracture temperature limits and for vessel thermal stresses during RPV heatup and cooldown. General Electric recommends that BWR owners/operators review this table against their current practices and evaluate any inconsistencies.

TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (Typical)

Measurement Use Limitations Steam dome saturation Primary measurement Must convert saturated temperature as determined above 212°F for Tech steam pressure to from main steam instrument Spec 100°F/hr heatup temperature.

line pressure and cooldown rate.

Recirc suction line Primary measurement Must have recirc flow.

coolant temperature. below 212OF for Tech Must comply with SIL 251 Spec IOOOF/hr heatup to avoid vessel stratification.

and cooldown rate.

Alternate measurement When above 212OF need to above 2120F. allow for temperature variations (up to 10-15OF lower than steam dome saturation temperature) caused primarily by FW flow variations.

D-2

GENE 0000-0000-8763-01a Revision 0 TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Alternate measurement for RPV drain line temperature (can use to comply with delta T limit between steam dome saturation temperature and bottom head drain line temperature).

RHR heat exchanger Alternate measurement Must have previously inlet coolant for Tech Spec 1OOOF/hr correlated RHR inlet temperature cooldown rate when in coolant temperature shutdown cooling mode. versus RPV coolant temperature.

RPV drain line Primary measurement to Must have drain line coolant temperature comply with Tech Spec flow. Otherwise, delta T limit between lower than actual steam dome saturated temperature and higher temp and drain line delta T's will be indicated coolant temperature. Delta T limit is 100°F for BWR/6s and 1450F for earlier BWRs.

Primary measurement to Must have drain line comply with Tech Spec flow. Use to verify brittle fracture compliance with Tech limits during cooldown. Spec minimum metal temperature/reactor pressure curves (using drain line temperature to represent bottom head metal temperature).

Alternate information Must compensate for outside only measurement for metal temperature lag bottom head inside/ during heatup/cooldown.

outside metal surface Should have drain line flow.

temperatures.

D-3

GENE 0000-0000-8763-Ola Revision 0 TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Closure head flanges Primary measurement for Use for metal (not coolant) outside surface T/Cs BWR/6s to comply with temperature. Install Tech Spec brittle fracture temporary T/Cs for metal temperature limit alternate measurement, if for head boltup. required.

One of two primary measure ments for BWR/6s for hydro test.

RPV flange-to-shell Primary measurement for Use for metal (not coolant) junction outside BWRs earlier than 6s to temperature. Response surface T/Cs comply with Tech Spec faster than closure head brittle fracture metal flange T/Cs.

temperature limit for head boltup.

One of two primary Use RPV closure head flange measurements for BWRs outside surface as alternate earlier than 6s for measurement.

hydro test. Preferred in lieu of closure head flange T/Cs if available.

RPV shell outside Information only. Slow to respond to RPV surface T/Cs coolant changes. Not available on BWR/6s.

Top head outside Information only. Very slow to respond to RPV surface T/Cs coolant changes. Not avail able on BWR/6s.

D-4

GENE 0000-0000-8763-O1 a Revision 0 TABLE OF RPV TEMPERATURE MONITORING MEASUREMENTS (CONTINUED)

(Typical)

Measurement Use Limitations Bottom head outside 1 of 2 primary measurements Should verify that vessel surface T/Cs to comply with stratification is not Tech Spec brittle fracture present for vessel hydro.

metal temperature (see SIL No. 251).

limit for hydro test.

Primary measurement to Use during heatup to verify comply with Tech Spec compliance with Tech Spec brittle fracture metal metal temperature/reactor temperature limits pressure curves.

during heatup.

Note: RPV vendor specified metal T limits for vessel heatup and cooldown should be checked during initial plant startup tests when initial RPV vessel heatup and cooldown tests are run.

D-5

GENE 0000-0000-8763-O1 a Revision 0 Product

Reference:

B21 Nuclear Boiler Prepared By: A.C. Tsang Approved for Issue: Issued By:

B.H. Eldridge, Mgr. D.L. AlIred, Manager Service Information Customer Service Information and Analysis Notice:

SILs pertain only to GE BWRs. GE prepares SILs exclusively as a service to owners of GE BWRs. GE does not consider or evaluate the applicability, if any, of information contained in SILs to any plant or facility other than GE BWRs as designed and furnished by GE. Determination of applicability of information contained in any SIL to a specific GE BWR and implementation of recommended action are responsibilities of the owner of that GE BWR.SILs are part of GE s continuing service to GE BWR owners. Each GE BWR is operated by and is under the control of its owner. Such operation involves activities of which GE has no knowledge and over which GE has no control. Therefore, GE makes no warranty or representation expressed or implied with respect to the accuracy, completeness or usefulness of information contained in SILs. GE assumes no responsibility for liability or damage, which may result from the use of information contained in SILs.

D-6

GENE 0000-0000-8763-01 a Revision 0 Appendix E Determination of Upper Shelf Energy E-1

GENE 0000-0000-8763-01a Revision 0 Paragraph IV.B of 10CFR50 Appendix G [1] sets limits on the upper shelf energy (USE) of the beltline materials. The USE must be above 50 ft-lb at all times during plant operation, assumed here to be up to 32 EFPY. Calculations of 32 EFPY USE, using Reg. Guide 1.99, Rev. 2 methods, are summarized in Table E-1. The values for initial USE were obtained from [2]

for all materials.

The USE decrease prediction values from Reg. Guide 1.99 [3] were used for the beltline plates and welds in Table E-1. Based on the above results, the beltline materials will have USE values above 50 ft-lb at 32 EFPY, as required in 10CFR50 Appendix G [1]. The lowest USE predicted for 32 EFPY is 75 ft-lb (for plate Heat C2557-1).

Since USE requirements are met, irradiation effects are not severe enough to necessitate additional analyses. Sufficient data is available to establish that the actual 32 EFPY USE values will be above 50 ft-lb, demonstrating acceptability.

E-2

GENE 0000-0000-8763-O1 a Revision 0 Table E-1. Upper Shelf Energy Analysis for Perry Beltline Material Location Heat Z ~est Tenipi Initial U.S.E.- %Cu U.S.E.b I32 EFPY U.S.E.'

PLATES Shell Rini #2 22-1-1 C2557-1 70 84 0.060 11 75 22-1-2 B6270-1 30 94 0.060 11 84 22-1-3 A1155-1 50 114 0.060 11 101 WELDS Lower Intermediate BD. BE. BF 5P6214B 10 88 0.020 10.5 79 BD, BF 627260 30 104 0.060 15.5 88 BD, BE, BF 626677 40 90 0.010 8.5 82 BE 624063 10 105 0.030 12 92 E 627069 0 112 0.010 8.5 102 a Values obtained from Perry Nuclear Power Plant Units I & 2 Updated Safety Analysis Report for Unit 1, Section 5.3 b Values obtained from Figure 2 of RG 1.99, Rev. 2 for 32 EFPY 1/4T fluence of 2.9 x 108 n/cm2 C 32 EFPY Transverse USE = Initial Transverse USE

• f1 - (% Decrease USE/ 100)1 E-3

GENE 0000-0000-8763-01a Revision 0 Appendix E References

1. "Fracture Toughness Requirements", Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, December 1995.

2. Perry Nuclear Power Plant Units 1 & 2 Updated Safety Analysis Report for Unit 1, Section 5.3, Revision 10.

3. "Radiation Embrittlement of Reactor Vessel Materials", USNRC Regulatory Guide 1.99, Revision 2, May 1988.

4. L.J. Tilly, "Perry Unit I RPV Surveillance Materials Testing and Analysis", GE-San Jose, CA, November 1996 (GE-NE-B1301793-01, Rev. 0).

E-4

GENE 0000-0000-8763-01a Revision 0 Appendix F P-T Curves for 32 EFPY 200°F/hr Heatup/Cooldown F-1

GENE 0000-0000-8763-Ola Revision 0 Table F-1. P-T Curves at 32 EFPY for 200°Flhr Heatup/Cooldown Figure Numbers Table Numbers Curve Curve Description for Presentation for Presentation of the P-T Curves of the P-T Curves Curve B Bottom Head Limits Figure F-1 Table F-2 UDper Vessel Limits Figure F-2 Table F-2 Beltline Limits Figure F-3 Table F-2 Curve C Bottom Head Limits Figure F-4 Table F-2 Uoper Vessel Limits Figure F-5 Table F-2 Beltline Limits Figure F-6 Table F-2 F-2

GENE 0000-0000-8763-Ola Revision 0 1400 1300 1200 1100 t

1000 w

IL 900 INITIAL RTNDT VALUE IS 0

10'F FOR BOTTOM HEAD J

S800 IL' Co o 700 HEATUP/COOLDOWN RATE OF COOLANT

< 200°F/HR S600 Z

i 500

'U S4 00

'U 0:

300 200

-BOTTOM HEAD LIMITS 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (fF)

Figure F-1. Bottom Head P-T Curve for Core Not Critical [Curve B]

[2000 F/hr or less coolant heatup/cooldown]

F-3

GENE 0000-0000-8763-01a Revision 0 1400 1300 1200 1100

• " 1000 Uj S- 900 0 INITIAL RTNDT VALUE IS

-20*F FOR UPPER VESSEL U) 800 U) w 0 700 U HEATUP/COOLDOWN RATE OF COOLANT S6 0 0 < 200°F/HR S5 0 0 w

S4 0 0 w

L.

300 200 SUPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE (OF)

Figure F-2. Upper Vessel P-T Curve for Core Not Critical [Curve B]

[2000 F/hr or less coolant heatup/cooldown]

F-4

GENE 0000-0000-8763-01 a Revision 0 1400 1300 1200 INITIAL RTNDT VALUE IS

-30°F FOR BELTLINE 1100 1000 BELTLINE CURVE 4

'U ADJUSTED AS SHOWN:

z EFPY SHIFT (-F)

(L 0 900 32 108 U

(0 800 w

o 700 HEATUP/COOLDOWN RATE OF COOLANT S600 <200°F/HR

500 w

S400 uJ 300 200

-BELTLINE LIMITS 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure F-3. Beltline P-T Curve for Core Not Critical [Curve B]

[2000 Flhr or less coolant heatup/cooldown]

F-5

GENE 0000-0000-8763-01a Revision 0 1400 1300 1200 1100 1000 w

x C. 900 INITIAL RTndt VALUE IS 0

10°F FOR BOTTOM HEAD 0

w 800 w HEATUP/COOLDOWN RATE OF COOLANT 6700 < 200°F/HR Z

500 w

S400 300 200 100 0

0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE ('F)

Figure F-4. Bottom Head P-T Curve for Core Critical [Curve C]

[2000 Flhr or less coolant heatup/cooldown]

F-6

GENE 0000-0000-8763-01 a Revision 0 1400 1300 1200 1100 0

1000 w

900 G RIO INITIAL RTNDT VALUE IS

-J -20°F FOR UPPER VESSEL S800 w

O 700 HEATUP/COOLDOWN 0

w RATE OF COOLANT S6 0 0 _<200°FIHR 0

FAGREION7 S5 0 0 S4 w

00 LU It 300 200 SUPPER VESSEL LIMITS (Including 100 Flange and FW Nozzle Limits) 0 0 25 50 75 100 125 150 175 200 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure F-5. Upper Vessel P-T Curve for Core Critical [Curve C]

[2000 F/hr or less coolant heatup/cooldown]

F-7

GENE 0000-0000-8763-01 a Revision 0 1400 INITIAL RTNDT VALUES ARE 1300 -30'F FOR BELTLINE,

-20'F FOR UPPER VESSEL, 1200 AND 10°F FOR BOTTOM HEAD 1100 U

CL 1000 BELTLINE CURVE ILl ADJUSTED AS SHOWN:

x EFPY SHIFT ('F)

IL 0 900 32 108

-J U'

( 800 U' HEATUP/COOLDOWN RATE OF COOLANT O 700 <200°F/HR S6 00 3 500 S400 a.

300 200

- BELTLINE CURVE 100 0

0 25 50 75 100 125 150 175 200 225 250 MINIMUM REACTOR VESSEL METAL TEMPERATURE

(°F)

Figure F-6. Beltline P-T Curve for Core Critical [Curve C]

[2000 F/hr or less coolant heatup/cooldown]

F-8

GENE 0000-0000-8763-01 a Revision 0 Table F-2. Perry P-T Curve Values for 32 EFPY Required Coolant Temperatures at 200°F/hr FOR FIGURES F-1, F-2, F-3, F-4, F-5, and F-6 BOTTOM UPPER BOTTOM UPPER HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVEB CURVEB CURVEB CURVEC CURVEC CURVEC (PSIG) 7OF) (7F) (7F) (7F) 7o.

0 70.0 70.0 70.0 70.0 70.0 70.0 10 70.0 70.0 70.0 70.0 70.0 70.0 20 70.0 70.0 70.0 70.0 70.0 70.0 30 70.0 70.0 70.0 70.0 70.0 70.0 40 70.0 70.0 70.0 70.0 70.0 70.0 50 70.0 70.0 70.0 70.0 70.0 70.0 60 70.0 70.0 70.0 70.0 70.0 70.0 70 70.0 70.0 70.0 70.0 70.0 70.0 80 70.0 70.0 70.0 70.0 70.0 70.0 90 70.0 70.0 70.0 70.0 70.0 70.0 100 70.0 70.0 70.0 70.0 70.0 70.0 110 70.0 70.0 70.0 70.0 70.0 70.0 120 70.0 70.0 70.0 70.0 70.0 70.0 130 70.0 70.0 70.0 70.0 70.0 70.0 140 70.0 70.0 70.0 70.0 70.0 70.0 150 70.0 70.0 70.0 70.0 70.0 70.0 160 70.0 70.0 70.0 70.0 70.0 70.0 170 70.0 70.0 70.0 70.0 70.0 70.0 180 70.0 70.0 70.0 70.0 70.0 70.0 190 70.0 70.0 70.0 70.0 70.0 70.0 200 70.0 70.0 70.0 70.0 70.0 70.0 210 70.0 70.0 70.0 70.0 70.0 70.0 220 70.0 70.0 70.0 70.0 70.0 70.0 230 70.0 70.0 70.0 70.0 70.0 70.0 240 70.0 70.0 70.0 70.0 70.0 70.0 250 70.0 70.0 70.0 70.0 70.0 70.0 260 70.0 70.0 70.0 70.0 70.0 70.0 270 70.0 70.0 70.0 70.0 70.0 70.0 280 70.0 70.0 70.0 70.0 70.0 70.0 290 70.0 70.0 70.0 70.0 70.0 70.0 300 70.0 70.0 70.0 70.0 70.0 70.0 310 70.0 70.0 71.5 70.0 70.0 70.0 312.5 70.0 70.0 72.1 70.0 70.0 70.0 312.5 70.0 130.0 130.0 70.0 130.0 70.0 320 70.0 130.0 130.0 70.0 130.0 70.0 330 70.0 130.0 130.0 70.0 130.0 70.0 340 70.0 130.0 130.0 70.0 130.0 70.0 350 70.0 130.0 130.0 70.0 130.0 70.0 360 70.0 130.0 130.0 70.0 130.0 70.0 370 70.0 130.0 130.0 70.0 130.0 70.0 380 70.0 130.0 130.0 70.0 130.0 70.0 390 70.0 130.0 130.0 70.0 130.0 70.0 400 70.0 130.0 130.0 70.0 130.0 70.0 F-9

GENE 0000-0000-8763-01 a Revision 0 Table F Cont. Perry P-T Curve Values for 32 EFPY Required Coolant Temperatures at 200°F/hr FOR FIGURES F-1, F-2, F-3, F-4, F-5, and F-6 BOTTOM UPPER BOTTOM UPPER HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE B CURVE B CURVE B CURVE C CURVE C CURVE C 4PSIG 70.0 13F. 13F. 70.0 170. 170.

410 70.0 130.0 130.0 70.0 170.0 170.0 420 70.0 130.0 130.0 70.0 170.0 170.0 430 70.0 130.0 130.0 70.0 170.0 170.0 440 70.0 130.0 130.0 70.0 170.0 170.0 450 70.0 130.0 130.0 70.0 170.0 170.0 460 70.0 130.0 130.0 70.0 170.0 170.0 470 70.0 130.0 130.0 70.0 170.0 170.0 480 70.0 130.0 130.0 70.0 170.0 170.0 490 70.0 130.0 130.0 70.0 170.0 170.0 500 70.0 130.0 130.0 70.0 170.0 170.0 510 70.0 130.0 130.0 70.0 170.0 170.0 520 70.0 130.0 130.0 70.0 170.0 170.0 530 70.0 130.0 130.0 71.2 170.0 170.0 540 70.0 130.0 130.0 73.9 170.0 170.0 550 70.0 130.0 130.0 7469 170.0 170.0 560 70.0 130.0 130.0 76.7 170.0 170.0 570 70.0 130.0 130.0 78.4 170.0 170.0 580 70.0 130.0 130.0 80.0 170.0 170.0 590 70.0 130.0 130.0 81.6 170.0 170.0 600 70.0 130.0 130.0 83.2 170.0 170.0 610 70.0 130.0 130.0 84.7 170.0 170.0 620 70.0 130.0 130.0 86.1 170.0 170.0 630 70.0 130.0 130.0 87.5 170.0 170.0 640 70.0 130.0 130.0 88.9 170.0 170.0 650 70.0 130.0 130.0 90.2 170.0 170.0 660 70.0 130.0 130.0 91.5 170.0 170.0 670 70.0 130.0 130.0 92.8 170.0 170.0 680 70.0 130.0 130.0 94.1 170.0 170.0 690 70.0 130.0 130.0 95.3 170.0 170.0 700 70.0 130.0 130.0 96.4 170.0 170.0 710 70.0 130.0 130.0 97.6 170.0 170.0 720 70.0 130.0 130.0 98.7 170.0 170.0 730 70.0 130.0 130.0 99.8 170.0 170.0 740 70.0 130.0 130.0 100.9 170.0 170.0 750 70.0 130.0 130.0 102.0 170.0 170.0 760 70.0 130.0 130.3 103.0 170.0 170.3 770 70.0 130.0 131.1 104.0 170.0 171.1 780 70.0 130.0 131.8 105.0 170.0 171.8 790 70.0 130.0 132.6 106.0 170.0 172.6 800 70.0 130.0 133.3 106.9 170.0 173.3 810 70.0 130.0 134.0 107.9 170.0 174.0 F-I 0

GENE 0000-0000-8763-01a Revision 0 Table F Cont. Perry P-T Curve Values for 32 EFPY Required Coolant Temperatures at 200°F/hr FOR FIGURES F-1, F-2, F-3, F-4, F-5, and F-6 BOTTOM UPPER BOTTOM UPPER HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE B CURVEB CURVE B CURVE C CURVE C CURVE C (PSIG (F) L °F E °F) (OF) (OF) (°F) 820 70.0 130.0 134.7 108.8 170.0 174.7 830 70.0 130.0 135.4 109.7 170.0 175.4 840 70.6 130.0 136.1 110.6 170.0 176.1 850 71.4 130.0 136.8 111.4 170.0 176.8 860 72.3 130.0 137.5 112.3 170.0 177.5 870 73.1 130.0 138.1 113.1 170.0 178.1 880 74.0 130.0 138.8 114.0 170.0 178.8 890 74.8 130.0 139.4 114.8 170.0 179.4 900 75.6 130.0 140.1 115.6 170.0 180.1 910 76.4 130.0 140.7 116.4 170.0 180.7 920 77.1 130.0 141.3 117.1 170.0 181.3 930 77.9 130.0 141.9 117.9 170.0 181.9 940 78.7 130.0 142.5 118.7 170.0 182.5 950 79.4 130.0 143.1 119.4 170.0 183.1 960 80.1 130.0 143.7 120.1 170.0 183.7 970 80.9 130.0 144.3 120.9 170.0 184.3 980 81.6 130.0 144.9 121.6 170.0 184.9 990 82.3 130.0 145.4 122.3 170.0 185.4 1000 83.0 130.0 146.0 123.0 170.0 186.0 1010 83.6 130.0 146.6 123.6 170.0 186.6 1020 84.3 130.0 147.1 124.3 170.0 187.1 1030 85.0 130.0 147.7 125.0 170.0 187.7 1040 85.6 130.0 148.2 125.6 170.0 188.2 1050 86.3 130.0 148.7 126.3 170.0 188.7 1060 86.9 130.0 149.3 126.9 170.0 189.3 1070 87.5 130.0 149.8 127.5 170.0 189.8 1080 88.2 130.0 150.3 128.2 170.0 190.3 1090 88.8 130.0 150.8 128.8 170.0 190.8 1100 89.4 130.0 151.3 129.4 170.0 191.3 1105 89.7 130.0 151.6 129.7 170.0 191.6 1110 90.0 130.0 151.8 130.0 170.0 191.8 1120 90.6 130.0 152.3 130.6 170.0 192.3 1130 91.2 130.0 152.8 131.2 170.0 192.8 1140 91.7 130.0 153.3 131.7 170.0 193.3 1150 92.3 130.0 153.8 132.3 170.0 193.8 1160 92.9 130.0 154.3 132.9 170.0 194.3 1170 93.4 130.0 154.7 133.4 170.0 194.7 1180 94.0 130.0 155.2 134.0 170.0 195.2 1190 94.5 130.0 155.7 134.5 170.0 195.7 1200 95.1 130.0 156.1 135.1 170.0 196.1 1210 95.6 130.0 156.6 135.6 170.0 196.6 F-11

GENE 0000-0000-8763-01a Revision 0 Table F Cont. Perry Limiting P-T Curve Values for 32 EFPY Required Coolant Temperatures at 200'F/hr FOR FIGURES F-I, F-2, F-3, F-4, F-5, and F-6 BOTTOM UPPER BOTTOM UPPER HEAD VESSEL BELTLINE HEAD VESSEL BELTLINE PRESSURE CURVE B CURVE B CURVE B CURVE C CURVE C CURVE C (PSIG) (OF) (OF) (OF) (OF) (OF) (OF) 1220 96.2 130.0 157.1 136.2 170.0 197.1 1230 96.7 130.0 157.5 136.7 170.0 197.5 1240 97.2 130.0 157.9 137.2 170.0 197.9 1250 97.7 130.0 158.4 137.7 170.0 198.4 1260 98.2 130.0 158.8 138.2 170.0 198.8 1270 98.7 130.0 159.3 138.7 170.0 199.3 1280 99.2 130.0 159.7 139.2 170.0 199.7 1290 99.7 130.0 160.1 139.7 170.0 200.1 1300 100.2 130.0 160.5 140.2 170.0 200.5 1310 100.7 130.0 160.9 140.7 170.0 200.9 1320 101.2 130.0 161.4 141.2 170.0 201.4 1330 101.6 130.0 161.8 141.6 170.0 201.8 1340 102.1 130.0 162.2 142.1 170.0 202.2 1350 102.6 130.0 162.6 142.6 170.0 202.6 1360 103.0 130.0 163.0 143.0 170.0 203.0 1370 103.5 130.0 163.4 143.5 170.0 203.4 1380 103.9 130.0 163.8 143.9 170.0 203.8 1390 104.4 130.0 164.2 144.4 170.0 204.2 1400 104.8 130.0 164.6 144.8 170.0 204.6 F-1 2

GENE 0000-0000-8763-O1 a Revision 0 Appendix G Determination of Beltline Region and the Impact on Fracture Toughness G-1

GENE 0000-0000-8763-Ola Revision 0 10CFR50, Appendix G defines the beltline region of the reactor vessel as follows:

"The region of the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

Furthermore, 10CFR50, Appendix H states that a surveillance program is not required if it can be demonstrated that the peak fluence (E>1 MEV) in the vessel does not exceed 1.0E17 n/cm2 .

Therefore, if it can be shown that no nozzles are located where the peak neutron fluence is expected to exceed or equal to 1.0E17 n/cm 2 , then it can be concluded that all reactor vessel nozzles are outside the beltline region of the reactor vessel.

The following dimensions are obtained from the referenced drawings:

Shell # 2 - Top of Active Fuel (TAF):. 363.5" (from vessel 0) (Reference 1)

Shell # 2 - Bottom of Active Fuel (BAF): 213.5" (from vessel 0) (Reference 1)

Elevation of Centerline of Recirculation Inlet Nozzle N2 in Shell # 1: 176.5" (from vessel 0)

(Reference 2)

Elevation of Centerline of Instrumentation Nozzle N12 in Shell #2: 364" (from vessel 0)

(Reference 3)

Elevation of Girth Weld between Shell #1 and Shell #2: 207.69" (from vessel 0) (Reference 3)

Elevation of Girth Weld between Shell #2 and Shell #3: 379.69" (from vessel 0) (Reference 3)

Based on the axial flux profile calculated for a similar plant, the RPV flux level at approximately 20" below the BAF and approximately 25" above TAF dropped to less than 0.01 of the peak flux level at the same radius. The redefined beltline region therefore spans from 193.5" (from vessel 0) to 388.5" (from vessel 0). From Reference 2, it can be seen that the Recirculation Inlet (N2) nozzle is closest to the beltline region (the top of the N2 nozzle is at an elevation of 192" above vessel 0), therefore none of the nozzles are within the redefined beltline region of the reactor vessel. It may be noted that the N12 Instrumentation nozzle is located within the beltline region; however this is a stainless steel penetration component and, as specified in Table A-2 of this report, is a discontinuity that does not require fracture toughness evaluation.

G-2

GENE 0000-0000-8763-01 a Revision 0 The girth welds between Shell #1 and #2 and between Shell #2 and #3, however, must now be considered within the redefined beltline region. Based on the same axial profile, it was determined that, at the elevation of the girth weld between Shell #1 and #2, the relative flux drops to a value 0.03 times the peak vessel flux; this fluence at 32 EFPY is 8.6E16 and therefore does not require further evaluation. Similarly, at the elevation of the girth weld between Shell #2 and

1. 3, the relative flux drops to a value 0.105 times the peak vessel flux. This fluence at 32 EFPY is 3.0E17 and is therefore evaluated in the table below. Note that no ART was high enough to consider these materials as limiting and therefore are not included in Tables 4-4a and 4-4b.

Based on the above, it is concluded that none of the Perry reactor vessel nozzles are within the redefined beltline region. It is further concluded and demonstrated in Table 4-4 of this report that the girth welds in the redefined beltline region are not limiting with respect to the beltline shift.

Table G-1. Perry Shell #3 ART Values (32 EFPY)1 Shell #2 Thickress in inches = 6.00 Ratio Peak/Location = 1.00 32 EFPY Peak I.D. fluence = 4.1E+18 n/crnA2 32 EFPY Peak 1/4 T ftuenoe = 2.9E+18 32 EFPY Peak 1/4 T ftuence = 2.9E+18 n/crnA2 Shell #2 Vertical Welds Thickness in inches= 6.00 Ratio Peak/Location = 1.00 32 EFPY Peak I.D. fluence = 4.1E+18 n/rcm2 32 EFPY Peak 1/4 T fluence = 2.9E+1B 32 EFPY Peak 1/4 T fluence = 2.9E+18 n/cmnA2 Initial 1/4 T 32 EFPY 32 EFPY 32 EFPY COMPONENT HEAT OR HEAT/LOT %Cu %Ni CF RTmT Fluence A RTNoT at oa Margin Shift ART

'F ncrn2 °F 'F =F 'F PLATES:

Shell #3 (Mark 23-1) 02432-1 0.05 0.61 31 -10 3.OE+17 7 0 3 7 14 4 C2432-2 0.08 0.61 51 10 3.OE+17 11 0 6 11 22 32 C2453-1 0.08 0.63 51 0 3.OE+17 11 0 6 11 22 22 Girth Seam AC 07R458 Lot S403B27AG 0.040 0.97 54 -60 3.OE+17 12 0 6 12 24 -36 Inthe absence of coper data. the mean from Shel #1 and Shell #2 data (0.0621 plus two standard deviations (2 x 0.008) was used to arrive at a conservative value of 0.08. Since no data was available for Shell #3, assumred that copper, nickel, and RTNDT would be sihnlar to that in Shell #2. With a fluence of 3.0E17. the verhical welds are fterefore not the limitingnatenial G-3

GENE 0000-0000-8763-01a Revision 0 Appendix G References

1. Design Input Request - Perry P-T Curves, Chuck Wirtz (FirstEnergy), 11/20/01

2. CB&I Nuclear Company, Drawing #21 Revision 4, "#1 Shell Ring Assembly", Chicago, Illinois (GENE VPF# 3521-212)

3. CB&I Nuclear Company, Drawing #22 Revision 6 "#2 Shell Ring Assembly", Chicago, Illinois (GENE VPF# 3251-213)

G-4

GENE 0000-0000-8763-01 a Revision 0 Appendix H Determination of Peak Vessel Fluence H-I

GENE 0000-0000-8763-01 a Revision 0 DETERMINATION OF LEAD FACTOR AND PEAK VESSEL FLUX Peak vessel flux can be determined from either a direct flux calculation or a combination of measured surveillance capsule flux and the lead factor. The flux wires detect flux at the location of the surveillance capsule. The wires will reflect the power fluctuations associated with the operation of the plant. However, the flux wires are not at the location of peak vessel flux. A lead factor is required in order to relate the flux at the location of the wires to the peak flux. The lead factor is the ratio of the flux at the surveillance capsule to the peak flux at the vessel inside surface. The lead factor is a function of the core and vessel geometry and depends on the distributions of power density and coolant voids in the core. The lead factor was calculated for the Perry geometry, using core data from Cycle 5 to determine power shape and void distribution.

The methodology used for the neutron flux calculation is documented in a Licensing Topical Report (LTR) NEDO-32983-A [1], which was approved by the NRC for licensing applications in the Safety Evaluation Report [2]. In general, GE's methodology described in the LTR follows the intent of Regulatory Guide 1.190 [3] for neutron flux evaluation. This methodology is briefly discussed below.

Procedure The lead factor and the flux distribution for the RPV inside wall were determined by using a combination of two separate two-dimensional neutron transport calculations.

The first of these calculations is performed in an (R,0) geometry to establish the azimuthal and radial variation of flux in the vessel at the core midplane elevation. The second calculation is performed in an (R,Z) geometry to determine the relative variation of flux with elevation. The azimuthal and axial distribution results were combined to provide a simulation of the three-dimensional distribution of flux. The ratio of fluxes, or lead factor, between the surveillance capsule location and the peak vessel flux location was obtained from this distribution.

The flux calculations are performed with DORTG01V, which is a discrete ordinates code package based on CCC-543 TORT-DORT Version 2.8.14 issued by Oak H-2

GENE 0000-0000-8763-O1 a Revision 0 Ridge National Laboratory (ORNL) in 1984 [4]. DORTGO1 is a controlled version of DORT in the GE Engineering Computation Program (ECP) library. The use of DORT for vessel flux calculations has been endorsed in the Regulatory Guide 1.190.

H-3

GENE 0000-0000-8763-01a Revision 0 Results for Cycle 5 The calculated peak fast flux (E>1 MeV) at the RPV inside surface is 3.59x1 09 n/cm 2-s. The azimuthal flux distribution at the vessel inner surface is illustrated in Figure H-2. The peak flux is located at angles of 25.50 and its mirror symmetry at 64.50 from the RPV quadrant references (00, 90°, etc.). The effects of inelastic scattering by steel are clearly displayed in the figure, where the flux depression occurs in regions shadowed by metal components.

Axial flux variation at the RPV inside surface is shown in Figure H-3. The elevation of peak flux occurs at about 101 inches above the bottom of the active fuel (BAF). The ratio of peak flux to midplane flux is 1.103.

The capsule from which the surveillance samples and flux wires were retrieved was located at 30. The calculated fast flux (E>1 MeV) at the capsule center is 1.86x1 09 n/cm 2-s. Therefore, the calculated lead factor for the RPV inside surface is 1.86 / 3.59 = 0.52.

This lead factor is a product of three spatial factors - radial, azimuthal, and axial, reflecting the positional differences between the surveillance capsule and the peak vessel flux. This new lead factor is significantly higher than the 1996 calculation of 0.42 [6]. This is attributed to the use of the new GE methodology that models the jet pump components in the flux calculations. The presence of jet pump components at the 300 azimuth significantly reduces the peak vessel flux due to the shadowing effect of the steel in the jet pumps. On the other hand, the shadowing effect does not affect the H-4

GENE 0000-0000-8763-Ola Revision 0 capsule, which is at the 30 azimuth and is not shadowed by the jet pump. The net effect is an increase of the lead factor.

The transport calculation of surveillance capsule flux, 1.86x10 9 n/cm 2-s, is about 9% lower than the dosimetry result of 2.04x10 9 n/cm 2-s. This is considered to be good agreement in view of the significant uncertainties in the analytical model and in the experimental results. However, it should be noted that the lead factor is not as sensitive to these differences as the magnitude of the flux. A difference in vessel radius has little, if any, effect on the calculated lead factor, since the difference would affect both capsule radius and vessel radius and would not significantly alter the ratio of fluxes at the two locations.

For the vessel flux to be used in the 32 EFPY fluence evaluation, there are two options. The first option is to use the peak vessel flux directly from the flux calculation, which is 3.59x109 n/cm2 -s. The second option is a combination of measured capsule flux of 2.04x109 n/cm 2-s and the calculated lead factor of 0.52. The peak vessel flux can be derived as 2.04x10 9 / 0.52 = 3.92x10 9 n/cm 2-s.

Since the second option has higher vessel flux than the first option, vessel flux from this option will be conservatively used for the fluence evaluation.

The fracture toughness analysis is based on a 1/4T depth flaw in the beltline region, so the attenuation of the flux to that depth is considered. This attenuation is calculated according to Reg. Guide 1.99 requirements, as shown in the following section titled Estimate of 32 EFPY Fluence.

Flux Estimate for Uprated Cycle The magnitude of vessel flux for the uprated cycle is estimated based on comparing the core design data for an uprated cycle with those for the Cycle 5.

H-5

GENE 0000-0000-8763-01 a Revision 0 Principally, the power distribution of the dominant peripheral bundles, which have significant contribution to the flux distribution at the vessel, is compared.

For the peak vessel flux, the dominant bundle is (1=21, J=1) due to its proximity to the peak flux location. Other neighboring bundles such as (1=23, J=2) and (1=22, J=2) also have significant contribution to the peak vessel flux. So the power distribution of these three bundles will be used for the comparison.

Figure H-4 shows a comparison of axial power distribution for the dominant peripheral bundles for Cycle 5 and the uprated cycle. For the three dominant peripheral bundles, the axial average is 0.418 for Cycle 5 and 0.339 for the uprated cycle. This 19% reduction gives indication that the peak vessel flux for the uprated cycle should be at least 5% lower than that for the Cycle 5. The axial power shape also gives some indication that the axial flux shape at the vessel would be peaked near the core midplane elevation for the uprated cycle.

Based on this comparison, it is concluded that the peak vessel flux for the uprated cycle is bounded by 105% of the Cycle 5 peak vessel flux.

ESTIMATE OF 32 EFPY FLUENCE The peak fluence at the vessel inside surface (fsurf) at 32 EFPY (end of life based on a 40-year operation at 80% capacity factor) is determined from the sum of products of vessel flux and the EFPY for each cycle. Based on the results of the previous section, the peak vessel flux for the pre-uprated cycle is 3.92x10 9 n/cm 2-s, or a peak fluence of 1.24x1017 n/cm 2 for one year. For the uprated cycles, a 105% factor is used to scale up the pre-uprated flux.

The following table shows the fluence calculation:

H-6

GENE 0000-0000-8763-01a Revision 0 Time Period Cumulative Fluence EFPY for Period Cumulative EFPY Relative Power Through Cycle 5 6.81E+17 5.50 5.50 100%

Through Cycle 8A 1.14E+18 3.73 9.23 100%

Through Cycle 8B 1.22E+18 0.64 9.87 103.5%

to 22 EFPY 2.80E+18 12.13 22.00 105%

to 32 EFPY 4.10E+18 10.00 32.00 105%

The resulting 32 EFPY fluence value at the peak vessel inside surface is:

furf = 4.10x10 18 n/cm 2 .

The 114 T fluence (f) is calculated according to the Reg. Guide 1.99 [5] equation:

f = fsurf (e-0. 2 4 x), (Eq. H-1) where x = distance, in inches, to the 1/4 T depth. The vessel beltline lower-intermediate shell ring is 6.23 inches thick ordered, 6.0 inches minimum requirement. The corresponding depth, x, taken from the minimum required thickness is 1.5 inches.

Equation H-1 evaluated for this value of x gives the 1/4 T value of 32 EFPY fluence, f = 2.86x10 18 n/cm 2 for the lower-intermediate shell ring. The 22 EFPY peak vessel fluence at the 1/4T is 2.00x1018 n/cm 2 . The determination of the fluence for girth welds AB and AC are discussed in Appendix G of this report.

H-7

GENE 0000-0000-8763-01a Revision 0 Table H-1 Summary of Daily Power History Cycle Date On Date Off [Days On (ti) MWd(t) I Cycle I Date On Date Off IDays On (ti) I MWd(t) 1 6/6/86 7/28186 53 3774 12/1/93 12/2/93 4994 7/28/86 3/24/87 240 26628 12/14/93 12/31/93 59853 4128/87 6/30/87 64 80170 111/94 1/31/94 99512 8/21187 1/3/88 136 235097 2/1/94 2/5/94 13214 1/29/88 5/18/88 111 337679 5 7125/94 7/27194 752 5/30/88 9/16/88 110 309440 811/94 8/31/94 55697 9/29/88 10121/88 23 69722 9/1/94 9/30/94 105893 11/4/88 2/22/89 111 368666 10/1/94 10/31/94 110497 7/24189 9/7/90 4111 1345154 11/1/94 11/30/94 107324 1/4/91 1/31/91 28 89705 12/1/94 12/31/94 95245 2/1/91 2/28/91 28 91513 1/1/95 1/31/95 110080 3/1191 3/31/91 31 107279 2/1/95 2/28195 96620 4/1/91 4/2/91 2 3254 3/1/95 3/31/95 104382 4118/91 4/30/91 13 43474 4/1/95 4/17/95 60351 511/91 5/31/91 31 108915 4/27/95 4/30/95 4295 611/91 6/30/91 30 106233 5/1/95 5/31/95 109779 7/1/91 7/31/91 31 101815 6/1/95 6/30/95 106760 8/1191 8131/91 31 108924 7/1/95 7/31/95 110036 9/1/91 9/30/91 30 107107 8/1/95 8/31/95 108890 10/1/91 10/31/91 31 95057 9/2/95 9/30/95 74190 11/1/91 11/30191 30 105587 10/1/95 10/31/95 110673 12/1/91 12/22/91 22 75434 11/1/95 11/11/95 34008 115/92 1/31192 27 93158 11/19/95 11/30/95 38377 2/1/92 2/29/92 29 103585 12/1/95 12/31/95 101609 3/1/92 3/21/92 21 70454 1/1/96 1/27/96 75282 6/13/92 6/30/92 18 44040 7/1/92 7/31/92 31 107547 Total Days Sum 8/1/92 8/31/92 31 108450 2614 7.18E+06 9/1/92 9/30/92 30 90400 10/1/92 10/31/92 31 81910 11/1/92 11/30/92 30 102213 12/1/92 12/31/92 31 108955 For Cycle 2, specific on/off dates are not 111193 1/9193 9 28182 presented, simply the total number of days in 3/7/93 3/26/93 20 44396 the cycle.

612/93 6/30/93 29 94537 7/1193 719/93 9 29453 7/27/93 7/31/93 5 6001 8/1/93 8/31/93 31 99547 9/1/93 9/30/93 30 105938 10/1/93 10/3/93 3 6227 11/16/93 11/30/93 15 37155 H-8

GENE 0000-0000-8763-01 a Revision 0 Table H-2 Summary of Perry Irradiation Periods CYCLE Cycle Core Cycle Cumulative Thermal Exposure Heavy EFPY EFPY Power (MWDIST) Metal (MWt)

Mass (ST) 0 1497.0 152.0338 0.17 0.17 3579 1 7916.6 152.0338 0.92 1.09 3579 2 8863.9 152.1452 1.03 2.12 3579 3 9289.0 151.9524 1.08 3.20 3579 4A 4465.0 150.3972 0.51 3.71 3579 4B 4056.1 150.3946 0.47 4.18 3579 5 11573.7 148.68 1.32 5.50 3579 6 11719.1 144.96 1.30 6.80 3579 7 12252.3 145.62 1.36 8.16 3579 8A 9535.2 146.63 1.07 9.23 3579 8B 5959.5 146.61 0.64 9.87 3704 9 (Projected) 17670.0 148.27 1.91 11.78 3758 H-9

GENE 0000-0000-8763-01a Revision 0 Table H-3 Surveillance Capsule Flux and Fluence for Irradiation from Start-up to 1/27/96 (30 Azimuth Capsule at 5.5 EFPY)

Wire Averagea Average Full Power Fluxb Full Power Fluxc Fluence Fluencec (Element) dps/g Element Reaction Rate (n/cm 2 -s) (n/cm 2 -s) (n/cm 2 ) (n/cm 2 )

(at end of irradiation) [dps/nucleus (saturated)] E>1 MeV E>O.1 MeV E>1 MeV E>0.1 MeV Copper 1.55E04 5.38E-18 2.04E09 3.26E09 3.53E17 5.65E17 Iron 1.46E05 3.18E-16 2.04E09 3.26E09 3.53E17 5.65E17 a Obtained by R.D Reager and L.K. Kessler b Full power flux, based on thermal power of 3579 MWt c 1.6 times the E >1 MeV result Measured Flux vs. Theoretical Flux at 30 Azimuth Dosimeter and Capsule E>1 MeV EFPY* Measured Flux Theoretical Flux (n/cm 2 -s) (n/cm 2 -s) 1989 EOC1 Dosimeter [8] 1.09 1.7x10 9 not available 1996 EOC5 Flux Wires 5.5 2.04xl 09 1.86xl 09

• Effective Full Power Years at 3579 MWt H-10

GENE 0000-0000-8763-01 a Revision 0 Figure H-1. Schematic of (R,O) Vessel Model 0U P 300

• Capsule/

Shroud 450 0

S E]I-1 I EI 1: n 0sDII i-I 0

3 L] I -I1-1 -1 I 00 4 -I [:] [:] [- -] D- -] M EIIII DLII]

6 EWDlD:1D 1:--lE1

-1~ F El/PIx

I1 DL 8 lID[D][El-:ID ID I 0

l-]W I DI- ED] I 11W I LIL ELELIELILILIEI 12 EE] LI ElIElILIDDElElDl1:

14 FD DF

-EN 15 17 181920 2I 2DD3IL El I25 2289:

El L El 900 1I-"'16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 H-11

GENE 0000-0000-8763-01 a Revision 0 Figure H-2. Azimuthal Distribution of Fast Neutron Flux at Vessel Inner Surface at Peak Elevation 4.OE+09 3.5E+09 3.OE+09 E

3 2.5E+09 L.

C 2.OE+09 P

1.5E+09 Z

1.OE+09 5.OE+08 0.0E+00 0 10 20 30 40 50 60 70 80 90 Azimuth (degrees past Quadrant Reference)

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GENE 0000-0000-8763-01 a Revision 0 Figure H-3. Axial Distribution of Relative Fast Neutron Flux at Vessel Inner Surface 1.0 0.9 I

0.8 0.7 0.6 X

0.5 0.4 0.3 0.2 0.1 0.0 0 25 50 75 100 125 150 Elevation Above BAF (Inches)

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GENE 0000-0000-8763-Ol a Revision 0 Figure H-4. Axial Power Distribution of Dominant Peripheral Bundles 0.6 0.5 0.4 0

0 .

0.3 0.2 0.1 0.0 0 25 50 75 100 125 150 Axial Elevation Above BAF (inches)

H-14

GENE 0000-0000-8763-01 a Revision 0 Appendix H

References:

1. NEDO-32983-A, Rev. 0, "Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations," December 2001.

2. Letter, S.A. Richards, USNRC to J.F. Klapproth, GE-NE, "Safety Evaluation for NEDC 32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC NO. MA9891)," MFN 01 -050, September 14, 2001.

3. Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," USNRC, March 2001.

4. CCC-543, "TORT-DORT Two- and Three-Dimensional Discrete Ordinates Transport Version 2.8.14," Radiation Shielding Information Center, Oak Ridge National Laboratory.

5. "Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988.

6. L.J. Tilly, "Perry Unit 1 RPV Surveillance Materials Testing and Analysis," GE-NE, San Jose, CA, November 1996, (GE-NE-B1301793-01, Rev. 0).

H-15